ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Idaho National Engineering and 

Environmental Laboratory (INEEL) – Occupational Environmental 

Dose

Document Number:

ORAUT-TKBS-0007-4 

Effective Date:03/30/2004

Revision No.:00

Controlled Copy No.: ______

Page 1 of 40

Subject Expert: Henry Peterson

Document Owner

Approval: Signature on File Date: 03/26/2004 Norman Rohrig, TBD Team Leader

Approval: Signature on File Date: 03/29/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 03/292004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 03/30/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................4

4.1 Introduction................................................................................................................................. 6

4.2 Internal Intakes From Onsite Airborne Radionuclide Concentration ............................................9

4.2.1 Operational Releases.......................................................................................................10

4.2.2 Episodic Releases at INEEL.............................................................................................12

4.2.2.1 SL-1 Reactor Accident.............................................................................................12

4.2.2.2 Criticalities ...............................................................................................................12

4.2.2.2.1 ICPP Criticality of October 16, 1959 ..............................................................12

4.2.2.2.2 ICPP Criticality of January 25, 1961 ..............................................................15

4.2.2.2.3 ICPP Criticality of October 17, 1978 ..............................................................16

4.2.2.2.4 HTRE No. 3 Criticality Excursion (IET 13)......................................................16

4.2.2.3 Releases from Planned Tests..................................................................................16

4.2.2.3.1 Fuel Element Burn Test A..............................................................................17

4.2.2.3.2 Fuel Element Burn Test B..............................................................................17

4.2.2.3.3 Fuel Element Cutting Facility Filter Break ......................................................17

4.2.2.3.4 Initial Engine Test 14 .....................................................................................18

4.2.2.3.5 Initial Engine Test 15(B) ................................................................................18

4.2.2.3.6 Initial Engine Test 16 .....................................................................................19

4.2.2.3.7 Initial Engine Test 17(B) ................................................................................19

4.2.2.3.8 Initial Engine Test 18 .....................................................................................19

4.2.2.3.9 Initial Engine Test 19(A) ................................................................................20

4.2.2.3.10 Initial Engine Test 25(A) ................................................................................21

4.2.2.3.11 Initial Engine Test 25(B) ................................................................................21
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4.2.2.3.12 Initial Engine Test 26(A) ................................................................................22

4.3 External Dose ...........................................................................................................................22

4.3.1 Facility Fence-Line Annual Doses ....................................................................................22

4.3.2 Facility Air Immersion Doses ............................................................................................24

4.4 Uncertainty................................................................................................................................24

References ........................................................................................................................................38

LIST OF TABLES

Table

4-1 Intake (Bq yr-1) by Year for ANL 1952-2002.............................................................................

Page

26

4-2 Intake (Bq yr-1) by Year for ARA 1952-2002.............................................................................27

4-3 Intake (Bq yr-1) by Year for CFA 1952-2002.............................................................................28

4-4 Intake (Bq yr-1) by Year for ICPP 1952-2002............................................................................29

4-5 Intake (Bq yr-1) by Year for RWMC 1952-2002 ........................................................................30

4-6 Intake (Bq yr-1) by Year for SPERT 1952-2002........................................................................31

4-7 Intake (Bq yr-1) by Year for TAN 1952-2002.............................................................................32

4-8 Intake (Bq yr-1) by Year for TRA 1952-2002.............................................................................33

4-9 Comparison of Calculated Facility Intakes with Intakes from Environmental 

Monitoring Results...................................................................................................................34

4-10 Intakes (Bq/event) for Criticalities ............................................................................................35

4-11 Intakes (Bq/event) for Special Tests ........................................................................................35

4-12 Intake (Bq/event) for Initial Engine Tests .................................................................................36

4-13 INEEL facility fence direct gamma values (TLD – Background) ...............................................37

LIST OF FIGURES

Figure

4-1 Idaho National Engineering and Environmental Laboratory (INEEL) site map...........................

Page

7

4-2 Chronology of NRTS/INEL/INEEL Facilities and Programs .......................................................9

4-3 Nine-year (1974 - 1983) average mesoscale dispersion isopleths of air concentration............ 11

4-4 Onsite and Distant Particulate Beta Concentrations in Air.......................................................13

4-5 CFA Air Monitoring (Gross Beta) Data for 1976 ......................................................................14

4-6 Dispersion Coefficient Contours for the SL-1 Accident............................................................15

4-7 Example of On-site TLD Monitoring Data................................................................................23
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 9/26/2003 00-A New Technical Basis Document for the Idaho 

National Engineering and Environmental 

Laboratory – Occupationally Environmental Dose. 

Initiated by Norman Rohrig.

Draft 10/13/2003 00-B Incorporates comments from Noel Savignac and 

Cindy Bloom. Initiated by Norman Rohrig.

Draft 12/05/2003 00-C Incorporates comments from NIOSH, Craig Little, 

and Laura McDowell-Boyer. Initiated by Norman 

Rohrig.

Draft 1/16/2004 00-D Incorporates comments from Cindy Bloom. 

Initiated by Norman Rohrig.

Draft 2/16/2004 00-E Incorporates comments from NIOSH for 

concurrence. Initiated by Norman Rohrig.

03/30/2004 03/30/2004 00 First approved issue. Initiated by Norman Rohrig.
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 4 of 40

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANL-W Argonne National Laboratory West

ARA Auxiliary Reactor Area

ATR Advanced Test Reactor

BORAX Boiling Water Reactor Experiment

Bq Becquerel (1 disintegration per second)

CDC Centers for Disease Control and Prevention

CERT Controlled Environmental Release Test

CFA Central Facilities Area

Ci curie

CPP (Idaho) Chemical Processing Plant 

CSC Computer Science Center (a facility in Idaho Falls)

CTF Core Test Facility

DOE U.S. Department of Energy

EBR-I Experimental Breeder Reactor No. 1

EBR-II Experimental Breeder Reactor No. 2

EFS Experimental Field Station

EMDR Environmental Monitoring Data Report

EMR Environmental Monitoring Report

EOCR Experimental Organic Cooled Reactor

ERDA Energy Research and Development Administration

ESRF Environmental Science and Research Foundation

ETR Engineering Test Reactor

F Fast (solubility rate)

FEBT Fuel Element Burn Test 

FECF Fuel Element Cutting Facility

FP Fission Product

FPFRT Fission Product Field Release Test

GCRE Gas-Cooled Reactor Experiment

GE General Electric (Company)

GE-ANP General Electric-Advanced Nuclear Propulsion (Program)

H&S Health & Safety (management organization)

HTRE Heat Transfer Reactor Experiment

ICPP Idaho Chemical Processing Plant

IDO, ID Idaho Operations Office (of AEC)

IET Initial Engine Test

INEEL Idaho National Engineering and Environmental Laboratory

INEL Idaho National Engineering Laboratory

INELHDE Idaho National Engineering Laboratory Historical Dose Evaluation

INTEC Idaho Nuclear Technology and Engineering Center

IRC INEL Research Center (a facility in Idaho Falls)
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kW killoWatts

LOFT Loss of Fluid Test (Facility)

LPT Low-Power Test (Facility)

LPTF Low Power Test Facility

M Moderate (solubility rate)

ML-1 Mobile Low-power Reactor No. 1

mR milli Roentgen 

mrem millirem

MTR Materials Testing Reactor

MW megawatt

NCRP National Council on Radiation Protection and Measurements

NOAA National Oceanic and Atmospheric Administration

NRF Naval Reactor Facility

NRTS National Reactor Testing Station

OMRE Organic Moderated Reactor Experiment

PBF Power Burst Facility

PREPP Process Experimental Pilot Plant 

RAC Radiological Assessment Corporation

RaLa radioactive lanthanum

RM radioactive material

RSAC Radiological Safety Analysis Computer (program)

RWMC Radioactive Waste Management Complex

RWMIS Radioactive Waste Management Information Service

S Slow (solubility rate)

SL-1 Stationary Low-Power Reactor No. 1

SMC Specific Manufacturing Capability

STPF Shield Test Pool Facility

SPERT Special Power Excursion Reactor Test

TAN Test Area North

TBD Technical Basis Document

TLD Thermoluminescent Dosimeter

TRA Test Reactor Area

TREAT Transient Reactor Test

TSF Technical Support Facility

WERF Waste Experimental Reduction Facility

WRRTF Water Reactor Research Test Facility

ZPPR Zero Power Plutonium Reactor
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4.1 INTRODUCTION 

This technical basis document (TBD) addresses radioactive material (RM) releases from areas or 

facilities at the Idaho National Engineering and Environmental Laboratory (INEEL), formerly the 

National Reactor Testing Station (NRTS) and later the Idaho National Engineering Laboratory (INEL) 

that could affect employees at another facility. The releases discussed here have been divided into 

two components: (1) normal operational releases, and (2) episodic releases that generally are of short 

duration. These RM releases potentially represent unrecorded or missed doses, either as direct 

gamma or beta-gamma from immersion in the radioactive gaseous cloud, for those individuals who 

did not have personal dosimetry to record the dose, or as internal doses from RM inhalation.

This TBD also addresses direct gamma doses resulting from facility operations. In general, these 

doses, if not controlled by management, increase with time and create a facility background dose. At 

INEEL, these facility background doses were recorded by film badges infrequently and inconsistently 

before 1970 and by thermoluminescent dosimeters (TLDs) on a routine basis since 1972. These 

facility background doses, or facility fence-line doses, as they are sometimes called, are a nebulous 

indication of a dose that workers could receive if they inhabited the facility. INEEL facility fence-line 

doses (minus background) are presented for 11 locations. 

As outlined and discussed in Part 2 of this INEEL Site Profile, the INEEL Site was chosen by the U.S. 

Atomic Energy Commission (AEC) as an isolated location for testing various reactor concepts. The 

Site is isolated from the public in two important aspects: (1) it is remotely located from population 

centers, and (2) it is hydrologically isolated because no surface streams originate on and flow to an 

offsite location and no streams cross the Site. Although the Site sits above the large Snake River 

Aquifer that eventually surfaces and enters the Snake River in the Hagerman Valley area, the annual 

flow rate of the water in the aquifer is measured at 5 to 15 feet per day. 

During the 50-year history of the site about 50 different reactor concepts have been designed, built, 

and operated at INEEL. All of these reactors have been prototype, low-power critical, or test reactors; 

no weapons production or commercial power reactors have been operated at INEEL. Most, if not all, 

of these reactors have used highly enriched (93% or higher) uranium as fuel. Only a few have 

produced significant airborne effluent: (1) the Heat Transfer Reactor Experiment (HTRE) reactors, 

operated under the General Electric-Aircraft Nuclear Propulsion (GE-ANP) Program at the north end 

of the Site at Test Area North (TAN), (2) test reactors [Materials Testing Reactor (MTR), Engineering 

Test Reactor (ETR), and Advanced Test Reactor (ATR)], all at the Test Reactor Area (TRA) near the 

middle southern end of the Site, and (3) the Experimental Breeder Reactor II (EBR-II), at Argonne 

National Laboratory-West (ANL-W) at the south-eastern corner of the Site. 

Another historically important airborne effluent producer is the Idaho Nuclear Technology and 

Engineering Center (INTEC), formerly known as the Idaho Chemical Processing Plant (ICPP). This 

facility, constructed in the early 1950s, began processing nuclear fuel in February 1953 and continued 

until 1992. Throughout its history, the Chem Plant, as it is commonly known, has reprocessed fuel 

from test reactors at INEEL, zirconium-clad fuel reclaimed from various reactors, stainless-clad fuel 

from EBR-II, and many AEC test reactors from around the world. Apart from the GE-ANP Program, 

which will be accounted for and discussed below, ICPP airborne releases have historically been the 

most radiologically significant releases at INEEL. Through the years that INEEL environmental 

monitoring reports have been published, ICPP airborne effluents have been attributed to creating the 

maximum INEEL boundary dose. Considering this fact, it should be suspected that ICPP airborne 

effluent would also be responsible for the maximum INEEL worker doses. Calculations performed for 

the INEEL TBD show that although ICPP airborne effluent is the most radiologically significant release 
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 7 of 40

at INEEL, the impact to workers is significantly below the allowable and acceptable limit. INEEL 

facility locations mentioned above are shown in Figure 4-1.

Figure 4-1. INEEL Site Map. 

From the beginning of operations at the INEEL Site, facility locations were chosen to limit the potential 

for operational releases at one facility to affect another facility. Because the Site encompasses 890 

square miles, there was ample room to place facilities with this principle in mind. Because the INEEL 

Site has an average elevation of 5,000 feet and is generally meteorologically characterized with a 

nocturnal inversion from the north-northeast and a daytime lapse condition with winds from the 

southwest, transitional weather regimes are less frequent than at lower elevations. The 50-year 

history of the Site has demonstrated that the large expanse of INEEL and this meteorological 

characteristic have been satisfactorily effective in maintaining the operational isolation of each facility.

Beginning with the GE-ANP program, which began in the early 1950s, the site has had the capability 

of plume tracking by aircraft. The local National Oceanic and Atmospheric Administration (NOAA) 

SOUTH 

SECURITY GATE

▪

▪ ▪ CERRO GRANDE

FRENCHMAN’S

 CABIN

▪

RENO RANCH
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field office, which was dedicated to Site needs and requirements, provided plume projection 

capabilities for various programs with a rather extensive network of meteorological monitoring stations 

(Yanskey, Markee, and Richter 1966). The plumes from all intentional planned releases [Controlled 

Environmental Release Test (CERT), Fission Product Field Release Test (FPFRT), Fuel Element 

Burn Tests A and B, etc.] were directed over an instrumented monitoring grid (GRID III), remotely 

located from other facilities, such that other onsite facilities were not affected by the release. In 

general, these tests were performed in support of a specific program (i.e., the FPFRT and the Fuel 

Element Burn Tests A and B were conducted to support the GE-ANP Program).

All of the airborne releases at INEEL that have occurred since the beginning of the Site were reviewed 

and analyzed as a result of a request from the U.S. Department of Energy (DOE). This request by the 

DOE Idaho Operations Office (IDO) was to evaluate the radiological impact to INEEL boundary 

individuals from airborne releases that had occurred since the beginning of operations at the Site. 

With the help of NOAA, which had hourly meteorological data from 1956 to that time, analyses were 

completed for all airborne releases that occurred at INEEL. The radiological consequences for an 

adult, a child, and an infant were calculated with Version 4 of the Radiological Safety Analysis 

Computer program RSAC-4 (Wenzel 1990). The results of the study were published in the Idaho 

National Engineering Laboratory Historical Dose Evaluation (DOE 1991); this TBD refers to that report 

as the INELHDE. All releases considered for that report are the basis for the releases considered in 

this TBD. In addition, all the releases documented in the INELHDE, operational and episodic, have 

been independently reviewed and found, with minor modifications, to be substantially appropriate. 

The review, conducted by the Radiological Assessment Corporation (RAC 2002) at the request of the 

Centers for Disease Control and Prevention (CDC) and the State of Idaho, also evaluated the 

methodology by which the RSAC-4 computer program performs dose calculations against the 

National Council on Radiological Protection and Measurements (NCRP) methodology. It stated: “As a 

final point, Tables 7, 8, 9a, 9b, 10a, and 10b, and Figures 18 and 19

Version 6 of the RSAC code (Wenzel and Schrader 2001) is used extensively in the current report to 

provide onsite concentrations due to episodic releases as well as other evaluations. For more 

information on the RSAC code, see Peterson (2004). 

confirm that the NCRP method 

was suitable for these ranking purposes when the results are compared with those using the RSAC 

code. In all cases, the RSAC code confirmed the results obtained using the NCRP methodology.” 

(RAC 2002, p. 57). 

Figure 4-2 shows the chronology of NRTS/INEL/INEEL facilities and programs. A few comments on 

the development of facilities at INEEL will clarify some of the questions that arise when viewing the 

INEEL map in Figure 4-1 and putting it into context with the timeline in Figure 4-2.

The Experimental Breeder Reactor No. 1 (EBR-I) was the first reactor to operate at the Site. It and 

the Boiling Water Reactor Experiments (BORAX-I through –V) were in the southwestern corner of 

INEEL, operated under the AEC Chicago Operations Office by the University of Chicago as Argonne 

National Laboratory- West. These essentially low-power reactors produced very little radioactive 

airborne effluent. As the EBR-I and BORAX programs were completed, ANL-W relocated to the 

eastern section of INEEL where EBR-II, the Transient Reactor Test (TREAT) Facility, the Zero Power 

Plutonium Reactor (ZPPR), etc., are presently located. The EBR-I location is now a historic landmark. 

The Stationary Low-Power Reactor (SL-1), the Mobile Low-Power Reactor (ML-1), and the GasCooled Reactor Experiment (GCRE), which were all operated for the U. S. Army, were at the Auxiliary 

Reactor Area (ARA). The NRTS Burial Ground became the Radioactive Waste Management 

Complex (RWMC). The short-lived Organic Moderated Reactor Experiment (OMRE) and 

Experimental Organic Cooled Reactor (EOCR) were at the EOCR location.
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Figure 4-2. Chronology of NRTS/INEL/INEEL Facilities and Programs. 

Essentially all of the major facility areas, the TRA, ICPP, TAN, Central Facilities Area (CFA), Naval 

Reactor Facility (NRF), Special Power Excursion Reactor Test (SPERT), and RWMC have operated 

since the early days of the Site. The major changes in these facility areas are the extent of operations 

at these facilities.

All inhaled quantities and concentrations referred to in this document apply to individuals stationed at 

the INEEL Site. DOE and INEEL contractor employees located principally in Idaho Falls, Idaho, in 

DOE/INEEL contractor facilities [Willow Creek Bldg., the INEL Research Center (IRC), the Computer 

Science Center (CSC), etc.] are not affected by Site activities and, thus, not subject to the inhaled 

quantities and concentrations.

4.2 INTERNAL INTAKES FROM ONSITE AIRBORNE RADIONUCLIDE CONCENTRATION

This section addresses onsite concentrations of radionuclides from normal operational releases and 

from shorter term releases such as those from criticality incidents that happened at ICPP and HTRE 

No. 3 during Initial Engine Test (IET) No. 13, Fuel Element Burn Tests, etc. As stated above, 

operational releases from ICPP and TRA have been the predominant and radiologically significant 
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releases at INEEL during the history of the Site. For more discussion of these releases and their 

relationship to other, less significant releases, see Peterson (2004), INELHDE (DOE 1991), or RAC 

2002. 

For worker dose reconstruction, the analyst should use default values for the calculation under 

consideration. When solubility is of concern, an insoluble oxide form for solids is recommended for 

analysis, with “S” and “M” type materials being the predominant form. Without more definitive 

information on the type of material, the dose reconstructor should use the material that maximizes the 

dose for a particular situation. When iodines are of concern, they should be considered “F” type 

materials.

4.2.1

To determine onsite concentrations of radionuclides from operational releases at INEEL facility 

locations, the same methodology used to determine offsite concentrations for annual environmental 

monitoring reports is used. The release for each year of operation is exactly the same as that 

documented in INELHDE (DOE 1991) with one exception: an analysis was performed to reduce the 

number of radionuclides and yet retain those that contributed about 95% of the inhalation dose. This 

analysis, included in Peterson (2004), reduced the number of radionuclides from 56 to 9 for the 

operational releases. 

Operational Releases

Meteorological dispersion factors applicable to each INEEL facility were picked from the annual 

average mesoscale dispersion isopleths of ground-level air concentrations as published in the 

environmental monitoring reports published for INEEL as described in INELHDE (DOE 1991, 

Appendix B). As described in that Appendix, dispersion isopleths are available for the years 

beginning in 1973, with the exception of 1978, when the telemetry system was being upgraded. For 

the years prior to 1973, a 9-year average of mesoscale dispersion isopleths of ground-level air 

concentrations (Bowman 1984), shown in Figure 4-3, was used1

. For 1993 to 2002, annual average 

mesoscale isopleths from the annual environmental reports (ESRF 1994, 1995, 1996, 1997, 1998, 

2000; Stoller 2002a,b,c) were used to calculate the facility annual concentration. 

Of the many facilities on INEEL, eight facility areas have been chosen for analysis: TAN, ICPP, TRA, 

RWMC, CFA, SPERT, ARA, and ANL-W. TAN includes IET, CTF, TSF, LOFT, SMC, WRRTF, 

PREPP, and LPTF, depending on the year of operation. The SPERT area includes WERF and PBF 

depending on the year of operation. Facilities such as Grid III and the Experimental Field Station 

(EFS) are inhabited infrequently and have not been included. These facilities were staffed with 

personnel who were normally employed at CFA. When an isopleth for a given year is chosen for a 

particular facility, such as SPERT, that isopleth is assumed to apply to PBF, SPERT-I, SPERT-II, etc. 

If a facility was between two isopleths, the higher-valued isopleth was chosen. Yearly isopleth values 

for each of the eight facilities have been extracted from the annual environmental monitoring reports 

and converted from the normalized annual concentration1

The annual inhaled quantities (Bq/yr) provided in Tables 4-1 through 4-8 for each of 8 facility areas 

are based on known and published INEEL annual airborne emissions. The following discussion 

provides information that could be located in NRTS/INEL/INEEL documentation concerning facility 

environmental sampling/monitoring and provides data that can be compared with these calculations. 

(hr2

/m3) to concentrations (Bq/m3) and 

multiplied by 2.4E+3 m3

/yr (the amount of air breathed occupationally per year) to produce activity 

inhaled per year (Bq) for an occupational individual. These are listed in Tables 4-1 through 4-8 for 

each of the facility areas. 

 

1 As used at INEEL, this quantity is the sum of 8,766 calculations of the hourly average χ/Q.
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Figure 4-3. Nine-year (1974 - 1983) average mesoscale dispersion isopleths of air 

concentration at ground level (hr2

/m3 x 10-9), normalized to unit release rate 

(Bowman 1984). 

“Environmental” air sampling at the facility areas has been performed at least since the mid 1950s 

and was accomplished where airborne effluents were known or suspected to exist. The early IDO 

H&S Division Annual Reports document many studies for defining radionuclide concentrations in the 

vicinity of different facilities. These studies were specific for a given test, operation, or incident 

however, and not performed in a set facility location, or for a standard time duration. Some facility 

data is presented in the 1963 Annual Progress Report of the IDO Health and Safety Division and a 

routine facility environmental monitoring program is developed between 1963 and 1970. In 1968 and 

1969, formal Environmental Monitoring Reports (EMRs) report alpha, beta, and I-131 concentrations 

that can be correlated with Table 4-1 through 4-8 values. The 1970 EMR reports gross beta values 

measured at CFA that can be correlated with Table 4-3 (CFA) values. EMRs between 1970 and 1990 

were reviewed for data that could be used for this correlation. Results of the comparison, showing 

that the calculated values of Tables 4-1 through 4-8 are in reasonable agreement with measured 

values reported by EMRs, are provided in Table 4-9. 

Figure 4-4 illustrates the variation of the INEEL Environmental Monitoring sampling results for the 9-

year period 1978 through 1986. This figure also illustrates the close correlation of environmental 

sample results acquired at “distant communities” and those acquired at the INEEL facilities and the 

effect of foreign nuclear tests and the Chernobyl reactor accident on INEEL environmental sampling 
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results. As shown on this figure for the 9-year period from 1978 through 1986, the INEEL average 

concentration has not differed from “distant community” concentrations by more than a factor of 2 for 

this 9-year period. Inspection of subsequent year’s Environmental Monitoring Reports shows the 

same is true for the years beyond 1986. 

It is also interesting to note that the greater perturbations in the facility, and distant community, 

concentrations are nearly all correlated with “fallout” from nuclear tests. An example of such a 

perturbation, attributed to a September 26, 1976, atmospheric test conducted by the People’s 

Republic of China, is illustrated in the graph of Figure 4-5 (ERDA 1976) where the normal 

concentration was increased by a factor of 20 and had a 3-month influence on the average 

concentration for all of the air concentrations measured, on-site and off-site. In the history of airmonitoring at the INEEL, an operational release has never approached the magnitude of perturbation 

created by fallout as illustrated in the graph of Figure 4-5. 

4.2.2

4.2.2.1 SL-1 Reactor Accident

Episodic Releases at INEEL

One significant accident at INEEL in the last 51 years released substantial amounts of radioactive 

material to the environment. The steam explosion at the SL-1 facility (near the present location of 

ARA II in Figure 4-1) killed three SL-1 personnel on January 3, 1961, and ruptured the SL-1 reactor 

vessel. This, in turn, propelled radioactive material into the reactor building and then into the 

environment. The amount of the release and the path that the cloud traveled from the reactor building 

was carefully monitored and well documented (Gammill 1961; Horan and Gammill 1961; Kunze 

1962). All radiological doses to personnel involved in the rescue and cleanup of the reactor building 

were carefully controlled and documented. 

The SL-1 accident did not affect any other INEEL facility with the effluent of radioactive material. The 

effluent traveled to the south of the facility, as shown in Figure 4-6 (DOE 1991).

4.2.2.2 Criticalities

Three accidental criticalities have occurred at the ICPP (now INTEC). The first occurred on October 

16, 1959, the second on January 25, 1961, and the third on October 17, 1978. The 1978 criticality 

released essentially just the noble fission gases produced during the criticality; halogens and solids 

were removed by the ICPP exhaust filtering system (Casto 1980). However, the two earlier 

criticalities did release radioactive material during or shortly after the event; in both cases, the effluent 

was transported to the south-southwest and potentially exposed personnel at the southern end of 

ICPP and at CFA. A conservative analysis, described below, defined the amount of potential 

radiological exposure that may have occurred to an individual at these locations. 

4.2.2.2.1 ICPP Criticality of October 16, 1959

On October 16, 1959, at approximately 3:00 a.m., the first criticality event occurred at the ICPP in the 

WH-100 vessel. The estimated magnitude of this event was no greater than 4 x 1019 fissions (DOE 

1991). Nuclear Incident at the Idaho Chemical Processing Plant (Ginkel et al. 1960) gives a full 

account of the incident and documents the radiological doses, calculated internal and measured 

external, for plant personnel involved in the incident.

For the calculation of intakes for this incident, meteorological conditions were modeled so that the Χ/Q 

at 22 km matched the value calculated for the Frenchman’s Cabin (located south of the INEEL as 
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Figure 4-4. Onsite and Distant Particulate Beta Concentrations in Air
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Figure 4-5: CFA Air Monitoring (Gross Beta) Data for 1976. 

 

 
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Figure 4-6. Dispersion Coefficient Contours for the SL-1 Accident (redrafted from Horan and Gammill, 

1961).

shown on Figure 4-1) where offsite doses were calculated and reported in DOE (1991). RSAC-6 was 

used to calculate Χ/Q values for the south end of the ICPP area and for the CFA. These 

concentrations and intake quantities would be applicable only if the individual was in the respective 

areas on the morning of October 16, 1959. The criticality occurred at approximately 3:00 a.m. Table 

4-10 lists the intakes applicable at the southern end of ICPP and CFA. 

4.2.2.2.2 ICPP Criticality of January 25, 1961

The January 25, 1961 criticality occurred in ICPP vessel H-110 about 9:50 a.m. This event consisted 

of an estimated 6.0 x 1017 fissions. The report documenting the incident states:

Of the 251 individuals present in the ICPP area at the time of the incident, none 

received significant radiation exposure. The highest exposure as determined from film 

badge readings did not exceed 55 millirem of penetrating radiation. Essentially no beta 

radiation was detected. No significant neutron exposure or internal contamination from

inhalation was found. The absence of significant exposures is attributable to the 

extensive shielding provided by the process cell in which the event took place and the 

control of the fission gases by the equipment. (Paulus et al 1961)
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As for the 1959 criticality, Χ/Q values were calculated for the southern end of the ICPP and for the 

CFA. The source term used for this event is the same as that used for DOE (1991). Table 4-10 lists 

the intakes applicable at the southern end of ICPP and CFA, if the individual was in the respective 

areas on January 25, 1961. 

4.2.2.2.3 ICPP Criticality of October 17, 1978

At approximately 8:40 p.m. on October 17, 1978, a criticality incident occurred in the first cycle 

uranium extraction system in the CPP-601 process building (DOE 1979; Casto 1980). The fissioning 

lasted for about 20 minutes and involved about 3.0 x 1018 fissions. This event did not result in 

significant radiation exposures to personnel, and there was no contamination of general plant areas. 

Releases to the environment which were filtered, consisted mainly of noble fission gases and small 

amounts of iodines. For this analysis, the release was treated as a 1-second release of noble fission 

gases and a small fraction of the halogens produced in the event. Table 4-10 lists the intakes 

applicable at the southern end of ICPP and CFA. These intake quantities would be applicable only if 

the individual was in the respective areas on the evening of October 17, 1978. 

4.2.2.2.4 HTRE No. 3 Criticality Excursion (IET 13)

IET 13 was characterized as the critical experiments and low-power testing phase of the HTRE No. 3 

reactor configuration for the GE-ANP Project. The low-power and critical experiments began on 

September 8 and ended on November 18, 1958, when a criticality excursion damaged every fuel 

element in the reactor core. As indicated in the GE literature concerning the test, the critical 

experiments and low-power testing of the assembly created insignificant radiological releases 

compared to the release resulting from the 770-megawatt-second energy excursion that ended this 

series of tests. More information concerning this incident is included in Peterson (2004). 

The released material for this test is consistent with the release modeled in INELHDE (DOE 1991). 

Table 4-10 provides the resulting inhaled quantities that would be applicable at the LPTF/STPF 

location. The intakes are applicable only if the individual were at the Low Power Test Facility/Shield 

Test Pool Facility (LPTF/STPF) location on November 18, 1958.

4.2.2.3 Releases from Planned Tests

Of the 108 episodic releases analyzed in DOE (1991), only 16 had the potential to affect other INEEL 

facilities. Section 4.2.2.2 describes three of these events, the criticalities at ICPP and the critical 

excursion in the HTRE No.3. The 12 remaining events are:

1. Fuel Element Burn Test A 7. IET 17(b)

2. Fuel Element Burn Test B 8. IET 18

3. Fuel Element Cutting Facility (FECF) 9. IET 19(A)

4. Initial Engine Test (IET) #14 10. IET 25(A)

5. IET 15(B) 11. IET 25(B)

6. IET 16 12. IET 26(A)

For a given test, if an onsite facility existed between the point at which the test was conducted and the 

affected offsite location, that test was conservatively assumed to have affected an onsite facility or 

facilities. For example, the meteorological conditions that existed during Fuel Element Burn Test A, at 

the Grid III location, dictated that the offsite location of Reno Ranch, adjacent to the northern INEEL 

boundary as shown on Figure 4-1, be evaluated in DOE (1991). Inspection of the plume passage 
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necessary to affect Reno Ranch indicates an impact at the TAN facility. The same conclusion for Fuel 

Element Burn Test B can be drawn, but in that case the plume impact on TAN facilities would be less 

direct because the offsite affected location was Birch Creek, which is considerably to the west of Reno 

Ranch. The Fuel Element Cutting Facility (FECF) Filter Break occurred at ICPP and clearly 

contaminated an area south of ICPP. According to the meteorological dispersion at the time of the 

filter break, the affected offsite location was Frenchman’s Cabin. Because CFA is in the straight-line 

path between ICPP and Frenchman’s Cabin, a radiological impact analysis was conducted for the 

CFA. 

Because all other test releases listed above, which originated at the TAN facility, affected two 

locations on the southern boundary [Frenchman’s Cabin or Cerro Grande (shown on Figure 4-1), as 

evaluated in DOE (1991)], they might also have affected ICPP, TRA, SPERT, south INEEL Security 

Gate, or CFA. Because TRA and ICPP are essentially the same distance from TAN, concentrations 

at either location were assumed to be the same. The following sections discuss these events.

4.2.2.3.1 Fuel Element Burn Test A

This test supported the GE-ANP program, providing information on the hypothetical radiological 

release during the crash of a nuclear-powered plane (Brodsky and Beard 1960). This test, with a 

stainless-steel-clad fuel element that had been operated at 20 MW for 120 hours and decayed for 70 

days, was conducted at Grid III at INEEL and began at 2:19 p.m. on March 20, 1957. The fuel 

element was assembled among other airframe parts to simulate an aircraft crash and burned for 2 

hours using jet fuel as the burning agent. After 2 hours, the fuel element, which was intact, was 

believed to have released not much radioactive material. Of the two Fuel Element Burn Tests, Burn 

Test A probably impacted TAN more directly than Burn Test B. Table 4-11 lists the best estimates of 

the radionuclides and TAN intakes of those nuclides. These intakes would have been received only if 

the individual was at TAN on March 20, 1957. 

4.2.2.3.2 Fuel Element Burn Test B

This test was conducted with a fuel element that had twice the curie content as the first burn test. The 

test began at 6:47 p.m. on the same day as the first burn test (March 20, 1957) and heated the fuel to 

much higher temperatures by supplying oxygen to the fire fueled by thermite, steel wool, and iron 

filings. The fuel element for this test underwent 6.16 x 1021 fissions and a subsequent decay of 250 

days (Brodsky and Beard 1960, p. 37). 

After 90 seconds, most of the fuel element had melted and dispersed. Figure 3-12 of DOE (1991) 

shows the trajectory of the plume, and Figure 3-13 shows the corresponding dispersion coefficient 

contours. Interpolation between the contours of 29 and 13,000 x 10-14 h m-3 provided a concentration 

for the GE-ANP area in the range of 6 x 10-12 h m-3, the value used for determining the dose of an 

individual at TAN. The radionuclides released and the respective curie values were the same as 

those for the DOE (1991) analysis. Table 4-11 lists the resulting intakes for TAN. These intakes 

would have been received only if the individual was at TAN on the evening of March 20, 1957. 

4.2.2.3.3 Fuel Element Cutting Facility Filter Break

Fuel elements sent to ICPP contained structural components on the ends that were cut off before the 

elements were processed. Cutting these end pieces off and cutting the fuel elements into sections 

before they were sent to CPP-601 for processing occurred in the FECF in CPP-603. During the night 

of October 29 and early in the morning of October 30, 1958, decontamination operations were 
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conducted in the FECF. Acid fumes from the decontamination operations caused failure of the FECF 

exhaust filters, resulting in the release of particulate activity to the south of ICPP. 

Approximately 100 curies (Ci) of long half-life particulate radioactive material was released over an 

area of approximately 200 acres (AEC 1959). The released radioactive material and quantities were 

the same as those published in DOE (1991). Table 4-11 lists the best-estimate intakes of the 

radionuclides in the southern end of ICPP and the CFA area. These intakes would be applicable only 

if the individual was in the area at the time of the release, i.e., during the night of October 29 and the 

early morning of October 30, 1958.

4.2.2.3.4 Initial Engine Test 14

IET 14 was the eighth nuclear test conducted by the GE-ANP program at TAN. This test was the fifth 

in the HTRE-2 reactor configuration. This test series involved the evaluation of the L2A-1 insert 

cartridge. The cartridge contained fueled and unfueled, beryllium-oxygen ceramic tubes. There was 

no coating on the inside surfaces of the fueled tubes (Pincock 1959).

A total of 100.25 hours was accumulated on the insert fuel cartridge at a maximum insert fuel 

temperature of approximately 2,500°

F. The objectives of the test were to (1) evaluate the operational 

effect of water vapor corrosion on fueled beryllium-oxygen tubes operating at a constant reactor 

mixed mean discharge air temperature over approximately 100 hours, and (2) measure the fission 

product release rate from uncoated fueled tubes as a function of temperature and operating time 

(Pincock 1959). 

Table 4-12 lists the of the fission products released during the IET #14 test, and the intakes at 

TRA/ICPP and CFA. An individual would have been exposed to these concentrations only if present 

at these locations between April 24 and May 19, 1959. The above radiological doses are for a total 

exposure period of 26 days.

4.2.2.3.5 Initial Engine Test 15(B) 

IET 15 was conducted at TAN between May 27 and June 24, 1959. This test involved the evaluation 

of the L2C-1 insert cartridge, which was of the concentric ring design. The fuel sheet was made of a 

chromium-uranium dioxide-titanium core clad with an iron-chromium-yttrium alloy (Evans 1959). From 

this operation data was obtained to evaluate:

1. Endurance capabilities of the advanced metals at a design temperature of 2,000°F for 

extended periods (planned endurance testing to total 120 hours or more)

2. The structural and metallurgical integrity of the fuel sheet in this particular cartridge design

3. The nature and extent of fuel sheet damage, if any, and the effect on cartridge performance

4. The performance potentials of the cartridge

The operation was successfully conducted to accumulate 80.75 hours at an insert extrapolated fuel 

sheet temperature of 2,015o

F. The operation was terminated after 80.75 hours due to a release of 

fission products of such a quantity as to indicate fuel sheet rupture of an extent sufficient to warrant 

inspection (Evans 1959).
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The insert was visually examined after completion of testing. No damage had occurred to the outer 

fuel sheets of the cartridge; however, blisters were observed on the inner fuel sheets. In some 

instances the blisters had ruptured. The FP release for this test was divided into two periods based 

on a review of effluent monitoring data. June 3 to 15 was considered to be an operation before the 

development of significant fuel sheet blisters. June 16 to 24 comprised the second period, when 

effects of blistering were clearly observed. 

According to the meteorology of the testing period, the second period affected Frenchman’s Cabin. 

Accordingly, this analysis addressed the radiological impact on the TRA/ICPP and CFA. Table 4-12 

lists the release of fission products, which correspond with the Part B operation release documented 

in DOE (1991) for the intakes applicable at TRA/ICPP and CFA. An individual would only have been 

exposed to these concentrations and intakes if present at the locations between June 16 and June 24, 

1959. The above radiological doses are for the total 9-day exposure period.

4.2.2.3.6 Initial Engine Test 16 

This was the first power test conducted in HTRE No. 3 to determine the operating characteristics of 

the horizontal core. Because the operation was to determine these characteristics, most operations 

were at low power levels. The total operation occurred between July 28 and October 9, 1959, and 

produced only 95 MW-hours of power (Showalter 1959). 

The fission product release for this test was modeled the same as that in DOE (1991) and was 

assumed to occur on October 9, 1959. Modeling releases for IET 16 involved the preservation of 

three factors: (1) a burnup of 95 MW-hours, (2) a conservative particulate release of 14 Ci (1.5 Ci h-1

for 9.5 hours), and (3) a conservative release fraction of 3.0 x 10-7 for iodine isotopes, the highest 

fraction measured during the test. To preserve these values and arrive at release fractions for other 

groups of radionuclides, engineering judgment and preliminary data from a few iterations of the RSAC 

program were used. To meet the stated criteria of 1.5 Ci h-1 for particulates, the noble gas release 

fraction was assumed to be 200 times greater than that for the iodines, and the release fraction for the 

solids was assumed to be 10% of the halogen release fraction. 

The release for this test series was modeled the same as that for DOE (1991) with Cerro Grande, 

south of the INEEL boundary, as the location of highest offsite concentration. For this analysis, a 

straight-line trajectory from TAN to Cerro Grande intercepted the SPERT facility and the INEEL south 

security gate located just north of the junction of highways 26 and 20 on the road to CFA as shown on 

Figure 4-1. Table 4-12 lists the intakes at the respective locations.

4.2.2.3.7 Initial Engine Test 17(B) 

IET 17 was performed between October 12 and December 12, 1959. Releases of airborne 

radioactivity occurred between November 2 and December 12, 1959, when the reactor was operated 

at power levels exceeding 100 kW. The test series involved the evaluation of the L2E-1 insert 

cartridge (Evans 1960). Table 4-12 lists the intakes. An individual would have received these intakes 

only if present at the respective locations between November 2 and December 12, 1959. The above 

radiological doses are for a total exposure period of 40 days.

4.2.2.3.8 Initial Engine Test 18

IET 18, conducted between December 23, 1959, and February 8, 1960, was designated as the Phase 

2 testing of the HTRE No. 3 engine and was an extension of the test program outlined for IET 16. The 

following is a brief description of operations during this test series:
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The powerplant was transported to Initial Engine Test on December 14, 1959, for final 

checkout in the facility prior to testing. The first engine operation was accomplished 

December 22, 1959, and the first data was taken December 23, 1959, (run No. I-1). 

The reactor was made critical on December 23, 1959, (run No. I-6). (Highberg et al. 

1960)

Damage was sustained to the instrumentation circuitry of the powerplant on January 6, 

1960, as a result of failure of the electric aftercooling blowers. The powerplant was 

returned to the Hot Shop on January 7, 1960, for repairs that were completed January

12, 1960. Testing resumed at Initial Engine Test on January 21, 1960, and finished on 

February 7, 1960.

The initial transfer from chemical operation to full nuclear operation occurred on 

January 26, 1960, (run No. 11-12), and the design conditions for endurance testing 

were initially attained at 11:58 p.m. on January 26, 1960 (run 11-32). A total of 126.42 

hours of operation was achieved at design conditions with a continuous operation of 

64.9 hours at these conditions. Operations were accomplished above 1% of design 

power for a total of 166.5 hours.

With respect to effluent monitoring:

Continuous effluent monitoring was maintained to measure and to record the activity 

released to the atmosphere by the powerplant. The maximum output was 8.6 

curies/hour (measured 10 minutes after release). The total output for the test series 

was 1157 curies (measured 10 minutes after release). The maximum release rate for 

I-131 was approximately 1.5 curies/hour (measured 10 minutes after release). The 

total offsite inhaled and ingested was below measurable amounts during this test 

series. (Highberg et al. 1960).

The release for this test series was modeled the same as DOE (1991) with Cerro Grande, south of the 

INEEL boundary, as the location of highest concentration. For this analysis, a straight-line trajectory 

from TAN to Cerro Grande intercepted the SPERT facility and the INEEL south security gate. Table 

4-12 lists the intakes for the two onsite locations. 

4.2.2.3.9 Initial Engine Test 19(A) 

IET 19, conducted between February 9 and April 30, 1960, was a test series in the HTRE No. 2 

reactor to evaluate the L2E-3 insert, which contained fueled and unfueled hexagonal beryllium-oxygen 

ceramic tubes. The tubes were coated on the inside with coextruded zirconia (zirconium dioxide) 

(Pincock 1960). The primary purposes for running the test were to:

1. Operate the L2E-3 fuel cartridge at peak temperatures of 2,500° F and 2,600° F for 100 hours 

or more at each temperature level to evaluate the effectiveness of the zirconium-dioxide 

coating against hydrolysis and the release of fission products

2. Operate the insert fuel cartridge at various temperature levels at specified intervals during the 

endurance testing to determine fission product release as a function of insert temperature

3. Obtain additional information on the effectiveness of an electrostatic precipitator in removing 

fission products from the reactor effluent (Pincock 1960).
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Pincock (1960) summarized the estimated total fission product release for the test runs based on spot 

sampling and reported them as 10-minute-decayed curies. The total fission product release reported 

for IET 19 was 2,892 Ci. The release for this test was modeled as for DOE (1991). Table 4-12 lists 

intakes for TRA/ICPP and CFA. An individual would have been exposed to these intakes only if 

present at the respective locations between February 17 and February 29, 1960. The above 

radiological doses are for the total exposure period of 13 days.

4.2.2.3.10 Initial Engine Test 25(A) 

IET 25, performed between November 15 and December 19, 1960, was an extension of the Phase II 

testing program conducted in IET 18. The test was conducted in the HTRE No. 3 reactor 

configuration. Releases of airborne radioactivity corresponding to the significant periods of operation 

were assumed to have occurred between November 22 and December 15, 1960. The release at IET 

25(A) was assumed to have occurred from November 22 through November 30, 1960.

The purposes of test series IET No. 25 were to demonstrate the capabilities of the fuel 

elements above design temperatures and to confirm that the powerplant could achieve 

a full nuclear start as predicted. The reactor went critical on November 14, 1960, and 

the test program was completed on December 19, 1960. (Linn 1962).

Only the following summary of effluent monitoring activities and results was available: 

“Continuous effluent monitoring was maintained to measure and record the activity 

released to the atmosphere by the powerplant. The maximum output was 3.4 

curies/hour (measured 10 minutes after release). The total output for the test series 

was 218 curies (measured 10 minutes after release). The maximum release rate for 

I-131 was approximately 0.7 curies/hour (measured 10 minutes after release). The 

total offsite inhaled and ingested dose was below measurable amounts during this test 

series.” (Highberg et al. 1961)

For this analysis the release was modeled as in DOE (1991) as a straight-line trajectory such that the 

centerline plume affected TRA/ICPP and CFA. The intakes for this test are listed in Table 4-12. An 

individual would have received intakes only if present at the respective locations between November 

22 and December 15, 1960. The tabulated intakes are for a total exposure period of 24 days.

4.2.2.3.11 Initial Engine Test 25(B) 

IET 25, performed between November 15 and December 19, 1960, was an extension of the Phase II 

testing program conducted in IET 18 and the second part of the IET 25 test (Linn 1962; Highberg et 

al. 1961). The releases for the test correspond with the significant periods of operation with IET 25(B) 

releases occurring from December 1 to December 15, 1960. Of the total release for IET 25, 76% was 

assumed to have been released during the 25(B) operation.

The releases for this test were the same as those modeled for DOE (1991) with a straight-line 

trajectory from TAN to SPERT and the INEEL south security gate. Table 4-12 lists the intakes 

corresponding to the two onsite locations. An individual would have received intakes only if present at 

the locations between December 1 and December 15, 1960. The above radiological intakes are for 

the total exposure period of 16 days.
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4.2.2.3.12 Initial Engine Test 26(A) 

IET 26, conducted in HTRE No. 2, was performed between December 22, 1960, and March 31, 1961 

(Field 1961). Releases of airborne activity for the total test were assumed to have occurred between 

December 23, 1960, and March 30, 1961, when the reactor operated at power levels exceeding 120 

kW. Releases for the IET 26(A) operation occurred from December 23 to 28, 1960. The insert being 

tested was the L2E-6 cartridge, which consisted of fueled and non-fueled, ceramic beryllium-oxide 

hexagonal tubes coated on the inner surface with zirconium dioxide.

The airborne release model was consistent with the model of DOE (1991) with an assumed 

straight-line trajectory between TAN and TRA/ICPP and CFA. The intakes are listed in Table 4-12 for 

TRA/ICPP and CFA. An individual would have been exposed to these intakes only if present at the 

respective locations between December 23 and December 28, 1960. The tabulated intakes are for a 

total exposure period of 6 days.

4.3 EXTERNAL DOSE

External radiation dose at a facility can be created by direct radiation from two sources: direct 

beta/gamma radiation from the facility, or gaseous effluents released from the facility or from adjacent 

facilities. In general, direct beta/gamma radiation from the facility will increase with time because the 

general contamination of the area will increase. In addition, as a facility ages, radioactive sources 

tend to accumulate at the facility, causing the general background to increase with time. A 

responsible Health & Safety (H&S) organization will observe and curb such a trend to prevent 

personnel exposures from increasing unnecessarily. An excellent example of another aspect that 

can cause facility background to increase is the operational tests that were conducted at the IET area 

at TAN. During an IET, such as one of those described in Section 4.2.2.3 where fuel damage created 

a significant release to the environment, the 76-inch duct from the HTRE engine to the stack was a 

very significant radiation source. That is why operational facilities for that area were heavily shielded 

and personnel were required to be within one of these shielded facilities during a test. The following 

sections discuss facility fence-line film badge and thermoluminescent dosimeter (TLD) data which 

recorded doses from gaseous fission product releases that had the potential for personnel exposure. 

More information on these two subjects is provided in Peterson (2004). 

4.3.1

Before 1970, many film badge or TLD measurements were made inside the Site boundary. During 

the IET period at TAN (1956 to 1961), many film badges were placed along the highways that 

triangulated the IET area and along some of the highways at the southern end of the Site. Initially, the 

badges were retrieved and read once a month. The frequency increased to six weeks in 1962 and

then changed back to monthly changes in 1963. Film badges were used up through 9 months of 

1966 and TLDs were used from that time. Beginning in 1967 the TLDs were changed on a semiannual basis. Recorded significant readings during the film badge period showed that the maximum 

badge reading increased by only a factor of 2 or 3 above background. However, the location of the 

badge with the increased reading was not identified. More information and film badge data for this 

early period are provided in Peterson (2004). The “detection limit” for the film badge reading was 

often quoted as 10 mrem for both a beta reading and a gamma reading (AEC 1963) and was quoted 

as being 10 mrem for the TLD when it was first used. With the annual background radiation field at 

the INEEL before operations started measured at 100 to 150 mrem/year, the monthly value of 8 to 13 

mrem is at the detection limit of the film badge or the TLD. Therefore, the uncertainty for monthly 

changeouts is higher than for less frequent changeouts. 

Facility Fence-Line Annual Doses
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Between the latter part of 1970 and the latter part of 1972, facility fence monitoring and facility 

locations had been established. From 1972 through 1983, facility fence TLD measurements, made on 

a 6-month basis with 5 TLDs located at each facility position, are available in the Environmental 

Monitoring Data Reports (EMDRs) for the INEL. The imported table of Figure 4.7 (ERDA 1976) 

shows that uncertainty can vary from less than 10% to 20% for a given set of readings. At each of the 

34 monitoring locations shown on this table, there are normally 5 TLDs for a potential of 170 readings 

for a 6-month period. For this particular 2-year set of data, 1.5% of the 136 set of readings are 

assigned a 2 sigma uncertainty of 16 to 20% and 18.4% of the readings are assigned a 2 sigma

Figure 4-7 Example of On-site TLD Monitoring Data. 
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uncertainty of 11 to 15%. However, 80% of the 136 values ascribed for the 34 locations over the 2-

year period have a 2 sigma uncertainty of less than 10%. 

To supply facility values for the 1952 to 1972 period, the highest 6-month value from 4/72 to 4/73 for 

the respective facility was multiplied by 2 and applied to each year between 1952 and 1972. (It 

should be remembered that not all of the listed facilities commenced operation or even existed in 

1952.) Facility fence TLD measurements could not be located for 1984 through 1992, but for 1993 

and beyond, such facility fence-line measurements are again included in the INEL/INEEL EMRs. For 

the 1984 to 1992 period where TLD measurements are missing, reasonable extrapolations are used 

to provide the missing values. In addition, background TLD measurements, corresponding with the 

facility fence TLD measuring periods, were recorded in the EMRs. All reduced facility fence-line TLD 

data (facility fence-line data minus background) in the EMRs are listed in Table 4-13. A more detailed 

discussion of the data is included in Peterson (2004). 

4.3.2

INEEL facility air-immersion (beta-gamma) doses could be calculated from the noble gas and halogen 

portions of the operational releases, and, if applicable, from the noble gas portion of the applicable 

episodic releases. This calculation should be unnecessary because these releases would be 

recorded in the fence-line TLD doses presented in Table 4-13. 

Facility Air Immersion Doses

4.4 UNCERTAINTY

A detailed discussion of the derivation of airborne releases for operational conditions and episodic 

events is provided in INELHDE (DOE 1991).

Operational Releases - Discussions with the authors of the INELHDE suggest that operational 

releases, which were monitored, could be low by a factor of not more than 2. When the annual 

normalized ground-level concentration values are applied to the operational releases, the uncertainty 

could be increased. However, in considering this increased uncertainty, it is interesting to note facility 

air monitoring results that are also discussed in the annual EMRs. In each case, the facility air 

concentration is compared to that concentration for a distant community, usually Idaho Falls. 

Normally the concentration is indistinguishable from the concentration for the distant community as 

discussed in Section 4.2.1. 

Episodic Releases - As described in INELHDE, the episodic releases are a maximum reasonable 

value, based on the amount of material available to be released, and the conditions of the respective 

test. For such a release, the inhaled quantities (in Bq) were maximized by assuming the downwind 

exposed individual was subjected to the plume centerline concentration for the total time, night and 

day in most cases, of the release. In spite of the original effort to be “reasonably conservative” in the 

exposure estimates, some of the authors, however, have stated that the release considered for a 

particular episodic event might be low by as much as a factor of 3.

A concerted effort has been made to reduce the number of radionuclides involved in the releases for 

the episodic events. Overall, the mix of radionuclides for all of the episodic events is complicated by 

the type (“aged” versus “fresh”), and the relative quantities of each. When viewed together, the 

episodic events can be categorized into three categories: a) criticalities that involve “fresh” fission 

products that have relatively short half lives when compared to radionuclides released from the Fuel 

Element Burn Tests, for example, b) releases involving long half-lived, aged fission products (Fuel 

Element Burn Tests and the release from the Fuel Element Cutting Facility Filter Break), and c) 

releases from the remaining IET tests that released short half-lived radionuclides, which are generally 
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characterized as “fresh” fission products, and long half-lived radionuclides, which are characterized as 

“aged” fission products. The latter category is unique to the GE-ANP Program because of the “directto-air” conversion nature of the tests. Therefore, within these categories, the number of radionuclides 

has been reduced to the number that preserves 95% of the original dose that was calculated for that 

particular location. 

During1988 a minor contamination incident occurred at the RWMC, which led to minor inhalation 

exposures to a few RWMC individuals, but that incident has not been incorporated into the 

environmental doses. The following paragraph, excerpted from the 1988 EMR (DOE 1989) 

summarizes site-monitoring information for this incident:

In early spring of 1988, waste boxes which had been stored on an outdoor asphalt pad at the 

RWMC (some since 1986) were to be moved indoors. A box which had been stored outside 

was discovered to be breached with some contamination spread. Further investigation of 

approximately 500 boxes on the pad found three other boxes that had been breached. Since 

these boxes had been stored on the pad for more than a year and were known to contain 

primarily Pu-238, it is now assumed that concentrations of that radionuclide reported on 

RWMC air filters in 1986 and 1988 were due to the small releases from those boxes. The 

highest RWMC concentration of Pu-238 in 1986 was 0.06% of the derived concentration guide 

and in 1988 was 0.04% of the guide. Cleanup and contamination controls were instituted at 

RWMC, and the entire stack of boxes was covered in November 1988 as an interim measure 

to control the spread of contamination prior to a permanent solution to the problem. 

Film Badge and TLD Measurements – As discussed in Section 4.3.1, above, the uncertainty 

of individual measurements, made with film badges and TLDs, can be as high as + 100%, 

depending on the frequency of changeout, i.e., once per month, which was generally the case 

with film badges. The data for 1952 to 1972 of Table 4-13 is based on the highest TLD 6-

month values of 1972 for the respective facility. Although the GE-ANP IET tests were 

conducted in the late 50s and early 60s (the last IET test, IET #26, ended on March 31, 1961), 

it should be remembered that tests with planned releases were administratively and 

meteorologically controlled so the airborne effluent traveled to the northeast over the 

monitoring grid such that adjacent facilities were not impacted. In spite of these controls, the 

1952 to 1962 values for the TAN areas (TAN-TSF, TAN-LOFT, and TAN-LPT) could be low by 

a factor of 3. However, after 1967 when facility fence-line measurements were routinely made 

with TLDs, with 5 TLDs at a given location, the uncertainty is generally ascribed at less than 

10%. Occasionally, less than about 20% of the time, these measurements have an ascribed 

level of uncertainty as high as 20%. 

Dose reconstruction for individuals whose location is unknown should use intakes provided by Table 

4-3 (CFA) and exposures for ICPP as provided in Table 4-13. The values suggested will maximize 

the resultant individual dose. 
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 26 of 40

Table 4-1: Intake (Bq yr-1) by Year for ANL 1952-2002. 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 5.5E+0 3.3E-2 0.0E+0 0.0E+0 0.0E+0 4.1E-1 0.0E+0 1.3E-1 5.0E-1

1953 5.5E+0 7.3E-2 1.3E+0 8.9E-4 1.3E-4 4.1E-1 4.0E-1 6.8E-1 8.6E-1

1954 1.4E+1 5.3E-2 3.3E+0 2.2E-3 3.3E-4 1.0E+0 9.9E-1 1.2E+0 2.1E+0

1955 1.9E+1 7.9E-2 4.7E+0 2.2E-3 3.3E-4 1.4E+0 1.4E+0 1.7E+0 3.0E+0

1956 2.2E+1 9.4E-1 5.4E+0 3.6E-3 5.3E-4 1.6E+0 1.7E+0 1.8E+0 3.5E+0

1957 4.4E+0 1.1E+2 1.1E+1 6.3E-3 9.3E-4 4.7E-1 1.6E+0 2.9E+0 1.7E+0

1958 6.0E+0 8.3E+1 1.6E+1 9.1E-3 1.3E-3 6.6E-1 1.2E+0 4.1E+0 1.3E+0

1959 4.8E+0 1.8E+1 1.3E+1 7.3E-3 1.1E-3 5.4E-1 6.1E-1 3.4E+0 6.7E-1

1960 6.6E-2 2.6E+0 1.4E-1 2.5E-3 3.7E-4 6.6E-3 7.6E-2 2.1E-1 8.2E-2

1961 4.6E-2 3.5E+0 3.3E-3 4.6E-4 6.8E-5 2.8E-3 2.0E-1 2.5E-1 2.1E-1

1962 2.1E-1 3.3E+0 4.6E-1 3.0E-4 4.5E-5 2.1E-2 2.0E-1 3.8E-1 2.2E-1

1963 3.3E+0 2.1E+0 9.4E+0 2.8E-3 4.2E-4 3.7E-1 1.3E-1 2.8E+0 1.4E-1

1964 1.8E+0 1.1E-1 0.0E+0 1.1E-4 1.6E-5 2.8E+1 3.0E-2 7.1E-1 2.0E+0

1965 4.6E+0 7.4E-1 0.0E+0 4.7E-3 7.0E-4 2.0E+0 0.0E+0 2.7E+0 1.8E+0

1966 2.8E+0 4.3E-1 0.0E+0 1.1E-3 1.6E-4 1.2E+1 0.0E+0 7.8E-1 1.2E+0

1967 7.0E-2 1.8E-1 0.0E+0 1.2E-4 1.8E-5 1.7E+0 0.0E+0 2.1E-1 6.6E-1

1968 5.0E+0 3.4E-1 0.0E+0 2.3E-3 3.4E-4 7.4E-1 0.0E+0 1.2E+0 6.4E-1

1969 2.8E-1 4.8E-1 0.0E+0 5.0E-4 7.4E-5 3.7E-1 0.0E+0 3.6E-1 5.3E-1

1970 6.6E-1 2.5E-5 0.0E+0 7.0E-4 1.1E-4 3.0E-1 0.0E+0 2.7E-1 5.5E-1

1971 2.5E+0 7.0E-1 0.0E+0 2.1E-3 3.1E-4 3.2E+0 0.0E+0 1.1E+0 4.7E-1

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1

1973 2.4E-2 1.4E-5 0.0E+0 2.4E-4 3.5E-5 1.1E-1 0.0E+0 6.2E-2 2.2E-2

1974 7.9E-3 1.3E-3 0.0E+0 1.1E-4 9.7E-6 4.4E-2 0.0E+0 3.7E-2 1.2E-1

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1

1976 5.2E-5 3.2E-5 0.0E+0 8.1E-6 3.6E-6 8.1E-4 0.0E+0 3.9E-4 3.0E-2

1977 2.0E-4 1.4E-4 0.0E+0 8.0E-5 3.4E-5 1.1E-2 0.0E+0 5.0E-3 4.3E-1

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2

1984 2.9E-4 9.7E-5 0.0E+0 1.9E-5 7.4E-6 3.2E-4 0.0E+0 1.3E-4 1.4E-2

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1

1986 2.9E-4 8.9E-5 0.0E+0 1.3E-6 9.7E-8 2.3E-3 0.0E+0 1.6E-5 4.3E-2

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1

1988 2.9E-4 1.4E-5 0.0E+0 1.1E-6 1.7E-7 1.5E-2 0.0E+0 2.7E-5 5.2E-1

1989 2.9E-4 9.7E-6 0.0E+0 5.7E-9 8.1E-10 1.6E-4 0.0E+0 7.4E-6 6.7E-2

1990 1.2E-4 3.8E-5 0.0E+0 9.4E-10 9.4E-10 4.2E-5 0.0E+0 2.0E-7 3.1E-2

1991 1.2E-4 1.6E-5 0.0E+0 1.0E-10 1.0E-10 5.2E-5 0.0E+0 9.7E-5 1.8E-2

1992 1.2E-4 1.8E-5 0.0E+0 1.7E-7 1.7E-7 1.4E-5 0.0E+0 8.3E-6 3.1E-2

1993 0.0E+0 3.8E-6 0.0E+0 0.0E+0 8.7E-11 3.5E-5 0.0E+0 3.8E-5 0.0E+0

1994 0.0E+0 3.1E-5 0.0E+0 0.0E+0 4.6E-8 0.0E+0 0.0E+0 8.0E-5 0.0E+0

1995 0.0E+0 2.1E-5 0.0E+0 3.3E-8 5.5E-9 0.0E+0 0.0E+0 2.3E-6 0.0E+0

1996 0.0E+0 2.8E-5 0.0E+0 2.2E-7 4.5E-9 0.0E+0 0.0E+0 1.1E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 1.8E-7 5.5E-8 0.0E+0 0.0E+0 2.4E-5 0.0E+0

1998 0.0E+0 2.3E-5 0.0E+0 1.7E-7 1.8E-8 0.0E+0 0.0E+0 1.1E-5 0.0E+0

1999 0.0E+0 3.1E-5 0.0E+0 7.5E-8 7.1E-9 0.0E+0 0.0E+0 4.4E-6 0.0E+0

2000 0.0E+0 1.9E-3 0.0E+0 3.6E-5 3.6E-7 0.0E+0 0.0E+0 3.5E-3 0.0E+0

2001 0.0E+0 1.0E-3 0.0E+0 2.4E-7 2.9E-5 0.0E+0 0.0E+0 1.2E-4 0.0E+0

2002 0.0E+0 1.6E-4 0.0E+0 6.6E-6 1.8E-5 0.0E+0 0.0E+0 3.8E-3 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 27 of 40

Table 4-2: Intake (Bq yr-1) by Year for ARA 1952-2002

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 7.9E+0 4.7E-2 0.0E+0 0.0E+0 0.0E+0 5.8E-1 0.0E+0 1.8E-1 7.2E-1 

1953 7.9E+0 1.0E-1 1.9E+0 1.3E-3 1.9E-4 5.8E-1 5.7E-1 9.7E-1 1.2E+0

1954 2.0E+1 7.6E-2 4.8E+0 3.2E-3 4.7E-4 1.5E+0 1.4E+0 1.7E+0 3.1E+0

1955 2.8E+1 1.1E-1 6.7E+0 3.2E-3 4.7E-4 2.0E+0 2.0E+0 2.4E+0 4.3E+0

1956 3.2E+1 1.3E+0 7.7E+0 5.1E-3 7.5E-4 2.3E+0 2.4E+0 2.5E+0 5.0E+0

1957 6.2E+0 1.6E+2 1.6E+1 9.0E-3 1.3E-3 6.8E-1 2.2E+0 4.1E+0 2.4E+0

1958 8.5E+0 1.2E+2 2.3E+1 1.3E-2 1.9E-3 9.4E-1 1.7E+0 5.9E+0 1.9E+0

1959 6.9E+0 2.6E+1 1.9E+1 1.0E-2 1.5E-3 7.7E-1 8.7E-1 4.9E+0 9.6E-1

1960 9.5E-2 3.8E+0 2.0E-1 3.6E-3 5.3E-4 9.4E-3 1.1E-1 3.0E-1 1.2E-1 

1961 6.6E-2 4.9E+0 4.7E-3 6.6E-4 9.7E-5 3.9E-3 2.8E-1 3.6E-1 3.1E-1

1962 3.0E-1 4.7E+0 6.5E-1 4.3E-4 6.4E-5 3.0E-2 2.9E-1 5.4E-1 3.1E-1 

1963 4.8E+0 3.1E+0 1.3E+1 4.1E-3 5.9E-4 5.4E-1 1.8E-1 4.0E+0 2.0E-1

1964 2.6E+0 1.6E-1 0.0E+0 1.6E-4 2.3E-5 3.9E+1 4.3E-2 1.0E+0 2.8E+0

1965 6.5E+0 1.1E+0 0.0E+0 6.8E-3 1.0E-3 2.8E+0 0.0E+0 3.9E+0 2.5E+0

1966 3.9E+0 6.2E-1 0.0E+0 1.5E-3 2.3E-4 1.8E+1 0.0E+0 1.1E+0 1.7E+0

1967 9.9E-2 2.6E-1 0.0E+0 1.7E-4 2.5E-5 2.5E+0 0.0E+0 3.0E-1 9.5E-1

1968 7.1E+0 4.8E-1 0.0E+0 3.3E-3 4.8E-4 1.1E+0 0.0E+0 1.7E+0 9.1E-1 

1969 4.0E-1 6.9E-1 0.0E+0 7.2E-4 1.1E-4 5.3E-1 0.0E+0 5.1E-1 7.6E-1

1970 9.5E-1 3.6E-5 0.0E+0 1.0E-3 1.5E-4 4.3E-1 0.0E+0 3.8E-1 7.8E-1 

1971 3.6E+0 9.9E-1 0.0E+0 3.0E-3 4.4E-4 4.6E+0 0.0E+0 1.6E+0 6.6E-1

1972 3.9E-1 4.2E-1 0.0E+0 9.2E-4 1.4E-4 7.0E-1 0.0E+0 4.0E-1 2.4E-1 

1973 8.1E-3 4.6E-6 0.0E+0 8.0E-5 1.2E-5 3.8E-2 0.0E+0 2.1E-2 7.3E-3

1974 2.4E-2 3.8E-3 0.0E+0 3.3E-4 2.9E-5 1.3E-1 0.0E+0 1.1E-1 3.7E-1 

1975 1.3E-2 6.5E-3 0.0E+0 1.8E-4 3.6E-5 9.1E-2 0.0E+0 2.8E-2 3.5E-1

1976 7.4E-5 4.6E-5 0.0E+0 1.2E-5 5.1E-6 1.2E-3 0.0E+0 5.5E-4 4.3E-2 

1977 2.9E-4 2.0E-4 0.0E+0 1.1E-4 4.8E-5 1.5E-2 0.0E+0 7.2E-3 6.1E-1

1978 5.4E-4 2.9E-3 0.0E+0 1.1E-4 1.1E-5 8.1E-3 0.0E+0 2.8E-3 5.0E-1 

1979 7.6E-4 4.2E-4 0.0E+0 2.0E-4 2.2E-5 5.5E-3 0.0E+0 3.8E-2 2.2E-1

1980 1.2E-3 6.2E-3 0.0E+0 1.3E-4 1.7E-5 2.7E-3 0.0E+0 1.8E-3 1.3E+0

1981 1.2E-3 1.6E-2 0.0E+0 2.6E-5 4.8E-6 2.6E-2 0.0E+0 1.4E-3 8.6E-1

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 

1983 4.1E-4 2.1E-3 0.0E+0 1.7E-4 2.3E-5 3.3E-3 0.0E+0 1.5E-4 5.2E-2

1984 4.1E-4 1.4E-4 0.0E+0 2.7E-5 1.1E-5 4.5E-4 0.0E+0 1.8E-4 2.1E-2 

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1

1986 1.2E-3 3.8E-4 0.0E+0 5.5E-6 4.2E-7 1.0E-2 0.0E+0 6.9E-5 1.9E-1 

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1

1988 2.9E-4 1.4E-5 0.0E+0 1.1E-6 1.7E-7 1.5E-2 0.0E+0 2.7E-5 5.2E-1 

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2

1992 4.1E-4 6.0E-5 0.0E+0 5.7E-7 5.7E-7 4.6E-5 0.0E+0 2.8E-5 1.0E-1 

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0

1995 0.0E+0 4.8E-5 0.0E+0 7.7E-8 1.3E-8 0.0E+0 0.0E+0 5.4E-6 0.0E+0

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0

1999 0.0E+0 1.0E-4 0.0E+0 2.5E-7 2.4E-8 0.0E+0 0.0E+0 1.5E-5 0.0E+0

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0

2002 0.0E+0 3.7E-4 0.0E+0 1.6E-5 4.2E-5 0.0E+0 0.0E+0 8.9E-3 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 28 of 40

Table 4-3: Intake (Bq yr-1) by Year for CFA 1952-2002. 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 1.7E+0 2.9E+0 3.7E+0

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 4.3E+0 5.1E+0 9.2E+0

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 6.0E+0 7.3E+0 1.3E+1

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 7.2E+0 7.6E+0 1.5E+1

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 6.6E+0 1.2E+1 7.3E+0

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 5.1E+0 1.8E+1 5.6E+0

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 2.6E+0 1.5E+1 2.9E+0

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 3.2E-1 9.0E-1 3.5E-1

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 8.5E-1 1.1E+0 9.2E-1

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 8.6E-1 1.6E+0 9.3E-1

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 5.4E-1 1.2E+1 6.0E-1

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 1.3E-1 3.0E+0 8.5E+0

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0

1968 2.1E+1 1.4E+0 0.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1

1973 2.4E-2 1.4E-5 0.0E+0 2.4E-4 3.5E-5 1.1E-1 0.0E+0 6.2E-2 2.2E-2

1974 7.9E-3 1.3E-3 0.0E+0 1.1E-4 9.7E-6 4.4E-2 0.0E+0 3.7E-2 1.2E-1

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1

1976 5.2E-5 3.2E-5 0.0E+0 8.1E-6 3.6E-6 8.1E-4 0.0E+0 3.9E-4 3.0E-2

1977 2.0E-4 1.4E-4 0.0E+0 8.0E-5 3.4E-5 1.1E-2 0.0E+0 5.0E-3 4.3E-1

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2

1984 2.9E-4 9.7E-5 0.0E+0 1.9E-5 7.4E-6 3.2E-4 0.0E+0 1.3E-4 1.4E-2

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1

1986 2.9E-4 8.9E-5 0.0E+0 1.3E-6 9.7E-8 2.3E-3 0.0E+0 1.6E-5 4.3E-2

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1

1988 2.9E-4 1.4E-5 0.0E+0 1.1E-6 1.7E-7 1.5E-2 0.0E+0 2.7E-5 5.2E-1

1989 2.9E-4 9.7E-6 0.0E+0 5.7E-9 8.1E-10 1.6E-4 0.0E+0 7.4E-6 6.7E-2

1990 1.2E-4 3.8E-5 0.0E+0 9.4E-10 9.4E-10 4.2E-5 0.0E+0 2.0E-7 3.1E-2

1991 1.2E-4 1.6E-5 0.0E+0 1.0E-10 1.0E-10 5.2E-5 0.0E+0 9.7E-5 1.8E-2

1992 1.2E-4 1.8E-5 0.0E+0 1.7E-7 1.7E-7 1.4E-5 0.0E+0 8.3E-6 3.1E-2

1993 0.0E+0 3.8E-6 0.0E+0 0.0E+0 8.7E-11 3.5E-5 0.0E+0 3.8E-5 0.0E+0

1994 0.0E+0 3.1E-5 0.0E+0 0.0E+0 4.6E-8 0.0E+0 0.0E+0 8.0E-5 0.0E+0

1995 0.0E+0 2.1E-5 0.0E+0 3.3E-8 5.5E-9 0.0E+0 0.0E+0 2.3E-6 0.0E+0

1996 0.0E+0 2.8E-5 0.0E+0 2.2E-7 4.5E-9 0.0E+0 0.0E+0 1.1E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 1.8E-7 5.5E-8 0.0E+0 0.0E+0 2.4E-5 0.0E+0

1998 0.0E+0 2.3E-5 0.0E+0 1.7E-7 1.8E-8 0.0E+0 0.0E+0 1.1E-5 0.0E+0

1999 0.0E+0 3.1E-5 0.0E+0 7.5E-8 7.1E-9 0.0E+0 0.0E+0 4.4E-6 0.0E+0

2000 0.0E+0 1.9E-3 0.0E+0 3.6E-5 3.6E-7 0.0E+0 0.0E+0 3.5E-3 0.0E+0

2001 0.0E+0 1.0E-3 0.0E+0 2.4E-7 2.9E-5 0.0E+0 0.0E+0 1.2E-4 0.0E+0

2002 0.0E+0 5.3E-4 0.0E+0 2.2E-5 6.0E-5 0.0E+0 0.0E+0 1.3E-2 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 29 of 40

Table 4-4: Intake (Bq yr-1) by Year for ICPP 1952-2002. 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 4.0E-1 2.9E+0 3.7E+0

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 9.9E-1 5.1E+0 9.2E+0

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 1.4E+0 7.3E+0 1.3E+1

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 1.7E+0 7.6E+0 1.5E+1

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 1.6E+0 1.2E+1 7.3E+0

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 1.2E+0 1.8E+1 5.6E+0

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 6.1E-1 1.5E+1 2.9E+0

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 7.6E-2 9.0E-1 3.5E-1

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 2.0E-1 1.1E+0 9.2E-1

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 2.0E-1 1.6E+0 9.3E-1

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 1.3E-1 1.2E+1 6.0E-1

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 3.0E-2 3.0E+0 8.5E+0

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0

1968 2.1E+1 1.4E+0 0.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0

1972 1.2E+0 1.2E+0 0.0E+0 2.8E-3 4.2E-4 2.1E+0 0.0E+0 1.2E+0 7.2E-1

1973 5.7E-2 3.2E-5 0.0E+0 5.6E-4 8.1E-5 2.7E-1 0.0E+0 1.5E-1 5.1E-2

1974 7.9E-2 1.3E-2 0.0E+0 1.1E-3 9.7E-5 4.4E-1 0.0E+0 3.7E-1 1.2E+0

1975 3.8E-2 1.9E-2 0.0E+0 5.5E-4 1.1E-4 2.7E-1 0.0E+0 8.3E-2 1.1E+0

1976 5.2E-4 3.2E-4 0.0E+0 8.1E-5 3.6E-5 8.1E-3 0.0E+0 3.9E-3 3.0E-1

1977 8.7E-4 5.9E-4 0.0E+0 3.4E-4 1.5E-4 4.5E-2 0.0E+0 2.1E-2 1.8E+0

1978 1.6E-3 8.7E-3 0.0E+0 3.2E-4 3.4E-5 2.4E-2 0.0E+0 8.3E-3 1.5E+0

1979 7.6E-4 4.2E-4 0.0E+0 2.0E-4 2.2E-5 5.5E-3 0.0E+0 3.8E-2 2.2E-1

1980 1.2E-3 6.2E-3 0.0E+0 1.3E-4 1.7E-5 2.7E-3 0.0E+0 1.8E-3 1.3E+0

1981 1.2E-3 1.6E-2 0.0E+0 2.6E-5 4.8E-6 2.6E-2 0.0E+0 1.4E-3 8.6E-1

1982 2.2E-4 6.7E-5 0.0E+0 2.1E-5 2.3E-6 6.3E-4 0.0E+0 4.0E-4 1.0E-1

1983 1.2E-3 6.2E-3 0.0E+0 5.2E-4 6.9E-5 1.0E-2 0.0E+0 4.5E-4 1.5E-1

1984 1.2E-3 4.2E-4 0.0E+0 8.0E-5 3.2E-5 1.4E-3 0.0E+0 5.5E-4 6.2E-2

1985 2.9E-3 2.1E-2 0.0E+0 5.3E-5 1.1E-5 2.3E-2 0.0E+0 1.5E-3 2.3E+0

1986 1.2E-3 3.8E-4 0.0E+0 5.5E-6 4.2E-7 1.0E-2 0.0E+0 6.9E-5 1.9E-1

1987 1.2E-3 1.8E-4 0.0E+0 5.9E-6 9.0E-7 1.3E-4 0.0E+0 9.7E-5 3.3E+0

1988 1.2E-3 5.9E-5 0.0E+0 4.8E-6 7.3E-7 6.6E-2 0.0E+0 1.2E-4 2.2E+0

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2

1992 1.2E-3 1.8E-4 0.0E+0 1.7E-6 1.7E-6 1.4E-4 0.0E+0 8.3E-5 3.1E-1 

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0

1995 0.0E+0 6.9E-5 0.0E+0 1.1E-7 1.8E-8 0.0E+0 0.0E+0 7.7E-6 0.0E+0

1996 0.0E+0 9.2E-5 0.0E+0 7.3E-7 1.5E-8 0.0E+0 0.0E+0 3.6E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 5.9E-7 1.8E-7 0.0E+0 0.0E+0 8.1E-5 0.0E+0

1998 0.0E+0 2.3E-4 0.0E+0 1.7E-6 1.8E-7 0.0E+0 0.0E+0 1.1E-4 0.0E+0

1999 0.0E+0 3.1E-4 0.0E+0 7.5E-7 7.1E-8 0.0E+0 0.0E+0 4.4E-5 0.0E+0

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0

2002 0.0E+0 1.6E-3 0.0E+0 6.7E-5 1.8E-4 0.0E+0 0.0E+0 3.8E-2 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 30 of 40

Table 4-5: Intake (Bq yr-1) by Year for RWMC 1952-2002. 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 5.5E+0 3.3E-2 0.0E+0 0.0E+0 0.0E+0 4.1E-1 0.0E+0 1.3E-1 5.0E-1 

1953 5.5E+0 7.3E-2 1.3E+0 8.9E-4 1.3E-4 4.1E-1 4.0E-1 6.8E-1 8.6E-1

1954 1.4E+1 5.3E-2 3.3E+0 2.2E-3 3.3E-4 1.0E+0 9.9E-1 1.2E+0 2.1E+0

1955 1.9E+1 7.9E-2 4.7E+0 2.2E-3 3.3E-4 1.4E+0 1.4E+0 1.7E+0 3.0E+0

1956 2.2E+1 9.4E-1 5.4E+0 3.6E-3 5.3E-4 1.6E+0 1.7E+0 1.8E+0 3.5E+0

1957 4.4E+0 1.1E+2 1.1E+1 6.3E-3 9.3E-4 4.7E-1 1.6E+0 2.9E+0 1.7E+0

1958 6.0E+0 8.3E+1 1.6E+1 9.1E-3 1.3E-3 6.6E-1 1.2E+0 4.1E+0 1.3E+0

1959 4.8E+0 1.8E+1 1.3E+1 7.3E-3 1.1E-3 5.4E-1 6.1E-1 3.4E+0 6.7E-1

1960 6.6E-2 2.6E+0 1.4E-1 2.5E-3 3.7E-4 6.6E-3 7.6E-2 2.1E-1 8.2E-2 

1961 4.6E-2 3.5E+0 3.3E-3 4.6E-4 6.8E-5 2.8E-3 2.0E-1 2.5E-1 2.1E-1

1962 2.1E-1 3.3E+0 4.6E-1 3.0E-4 4.5E-5 2.1E-2 2.0E-1 3.8E-1 2.2E-1 

1963 3.3E+0 2.1E+0 9.4E+0 2.8E-3 4.2E-4 3.7E-1 1.3E-1 2.8E+0 1.4E-1

1964 1.8E+0 1.1E-1 0.0E+0 1.1E-4 1.6E-5 2.8E+1 3.0E-2 7.1E-1 2.0E+0

1965 4.6E+0 7.4E-1 0.0E+0 4.7E-3 7.0E-4 2.0E+0 0.0E+0 2.7E+0 1.8E+0

1966 2.8E+0 4.3E-1 0.0E+0 1.1E-3 1.6E-4 1.2E+1 0.0E+0 7.8E-1 1.2E+0

1967 7.0E-2 1.8E-1 0.0E+0 1.2E-4 1.8E-5 1.7E+0 0.0E+0 2.1E-1 6.6E-1

1968 5.0E+0 3.4E-1 0.0E+0 2.3E-3 3.4E-4 7.4E-1 0.0E+0 1.2E+0 6.4E-1 

1969 2.8E-1 4.8E-1 0.0E+0 5.0E-4 7.4E-5 3.7E-1 0.0E+0 3.6E-1 5.3E-1

1970 6.6E-1 2.5E-5 0.0E+0 7.0E-4 1.1E-4 3.0E-1 0.0E+0 2.7E-1 5.5E-1 

1971 2.5E+0 7.0E-1 0.0E+0 2.1E-3 3.1E-4 3.2E+0 0.0E+0 1.1E+0 4.7E-1

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1 

1973 5.7E-2 3.2E-5 0.0E+0 5.6E-4 8.1E-5 2.7E-1 0.0E+0 1.5E-1 5.1E-2

1974 2.4E-2 3.8E-3 0.0E+0 3.3E-4 2.9E-5 1.3E-1 0.0E+0 1.1E-1 3.7E-1 

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1

1976 7.4E-5 4.6E-5 0.0E+0 1.2E-5 5.1E-6 1.2E-3 0.0E+0 5.5E-4 4.3E-2 

1977 2.0E-4 1.4E-4 0.0E+0 8.0E-5 3.4E-5 1.1E-2 0.0E+0 5.0E-3 4.3E-1

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1 

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1 

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2

1984 4.1E-4 1.4E-4 0.0E+0 2.7E-5 1.1E-5 4.5E-4 0.0E+0 1.8E-4 2.1E-2 

1985 4.1E-4 3.0E-3 0.0E+0 7.6E-6 1.5E-6 3.3E-3 0.0E+0 2.2E-4 3.3E-1

1986 4.1E-4 1.3E-4 0.0E+0 1.8E-6 1.4E-7 3.3E-3 0.0E+0 2.3E-5 6.2E-2 

1987 4.1E-4 6.1E-5 0.0E+0 2.0E-6 3.0E-7 4.3E-5 0.0E+0 3.2E-5 1.1E+0

1988 1.2E-3 5.9E-5 0.0E+0 4.8E-6 7.3E-7 6.6E-2 0.0E+0 1.2E-4 2.2E+0

1989 4.1E-4 1.4E-5 0.0E+0 8.2E-9 1.2E-9 2.3E-4 0.0E+0 1.1E-5 9.6E-2

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2

1992 4.1E-4 6.0E-5 0.0E+0 5.7E-7 5.7E-7 4.6E-5 0.0E+0 2.8E-5 1.0E-1 

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0

1995 0.0E+0 6.9E-5 0.0E+0 1.1E-7 1.8E-8 0.0E+0 0.0E+0 7.7E-6 0.0E+0

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0

1999 0.0E+0 7.2E-5 0.0E+0 1.8E-7 1.7E-8 0.0E+0 0.0E+0 1.0E-5 0.0E+0

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0

2002 0.0E+0 5.3E-4 0.0E+0 2.2E-5 6.0E-5 0.0E+0 0.0E+0 1.3E-2 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 31 of 40

Table 4-6: Intake (Bq yr-1) by Year for SPERT 1952-2002. 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 1.7E+0 2.9E+0 3.7E+0

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 4.3E+0 5.1E+0 9.2E+0

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 6.0E+0 7.3E+0 1.3E+1

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 7.2E+0 7.6E+0 1.5E+1

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 6.6E+0 1.2E+1 7.3E+0

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 5.1E+0 1.8E+1 5.6E+0

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 2.6E+0 1.5E+1 2.9E+0

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 3.2E-1 9.0E-1 3.5E-1 

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 8.5E-1 1.1E+0 9.2E-1

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 8.6E-1 1.6E+0 9.3E-1 

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 5.4E-1 1.2E+1 6.0E-1

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 1.3E-1 3.0E+0 8.5E+0

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0

1968 2.1E+1 1.4E+0 0.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0

1972 1.2E+0 1.2E+0 0.0E+0 2.8E-3 4.2E-4 2.1E+0 0.0E+0 1.2E+0 7.2E-1 

1973 8.1E-3 4.6E-6 0.0E+0 8.0E-5 1.2E-5 3.8E-2 0.0E+0 2.1E-2 7.3E-3

1974 5.5E-2 8.9E-3 0.0E+0 7.8E-4 6.8E-5 3.1E-1 0.0E+0 2.6E-1 8.6E-1 

1975 3.8E-2 1.9E-2 0.0E+0 5.5E-4 1.1E-4 2.7E-1 0.0E+0 8.3E-2 1.1E+0

1976 7.4E-4 4.6E-4 0.0E+0 1.2E-4 5.1E-5 1.2E-2 0.0E+0 5.5E-3 4.3E-1 

1977 8.7E-4 5.9E-4 0.0E+0 3.4E-4 1.5E-4 4.5E-2 0.0E+0 2.1E-2 1.8E+0

1978 1.6E-3 8.7E-3 0.0E+0 3.2E-4 3.4E-5 2.4E-2 0.0E+0 8.3E-3 1.5E+0

1979 2.5E-3 1.4E-3 0.0E+0 6.8E-4 7.4E-5 1.8E-2 0.0E+0 1.3E-1 7.3E-1

1980 1.2E-3 6.2E-3 0.0E+0 1.3E-4 1.7E-5 2.7E-3 0.0E+0 1.8E-3 1.3E+0

1981 1.2E-3 1.6E-2 0.0E+0 2.6E-5 4.8E-6 2.6E-2 0.0E+0 1.4E-3 8.6E-1

1982 2.2E-4 6.7E-5 0.0E+0 2.1E-5 2.3E-6 6.3E-4 0.0E+0 4.0E-4 1.0E-1 

1983 1.2E-3 6.2E-3 0.0E+0 5.2E-4 6.9E-5 1.0E-2 0.0E+0 4.5E-4 1.5E-1

1984 1.2E-3 4.2E-4 0.0E+0 8.0E-5 3.2E-5 1.4E-3 0.0E+0 5.5E-4 6.2E-2 

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1

1986 1.2E-3 3.8E-4 0.0E+0 5.5E-6 4.2E-7 1.0E-2 0.0E+0 6.9E-5 1.9E-1 

1987 1.2E-3 1.8E-4 0.0E+0 5.9E-6 9.0E-7 1.3E-4 0.0E+0 9.7E-5 3.3E+0

1988 4.1E-4 2.0E-5 0.0E+0 1.6E-6 2.4E-7 2.2E-2 0.0E+0 3.9E-5 7.4E-1 

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2

1992 1.2E-3 1.8E-4 0.0E+0 1.7E-6 1.7E-6 1.4E-4 0.0E+0 8.3E-5 3.1E-1 

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0

1995 0.0E+0 6.9E-5 0.0E+0 1.1E-7 1.8E-8 0.0E+0 0.0E+0 7.7E-6 0.0E+0

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0

1999 0.0E+0 3.1E-4 0.0E+0 7.5E-7 7.1E-8 0.0E+0 0.0E+0 4.4E-5 0.0E+0

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0

2002 0.0E+0 3.7E-4 0.0E+0 1.6E-5 4.2E-5 0.0E+0 0.0E+0 8.9E-3 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 32 of 40

Table 4-7. Intake (Bq yr-1) by Year for TAN 1952-2002. 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 5.5E+0 3.3E-2 0.0E+0 0.0E+0 0.0E+0 4.1E-1 0.0E+0 1.3E-1 5.0E-1 

1953 5.5E+0 7.3E-2 1.3E+0 8.9E-4 1.3E-4 4.1E-1 4.0E-1 6.8E-1 8.6E-1

1954 1.4E+1 5.3E-2 3.3E+0 2.2E-3 3.3E-4 1.0E+0 9.9E-1 1.2E+0 2.1E+0

1955 1.9E+1 7.9E-2 4.7E+0 2.2E-3 3.3E-4 1.4E+0 1.4E+0 1.7E+0 3.0E+0

1956 2.2E+1 9.4E-1 5.4E+0 3.6E-3 5.3E-4 1.6E+0 1.7E+0 1.8E+0 3.5E+0

1957 4.4E+0 1.1E+2 1.1E+1 6.3E-3 9.3E-4 4.7E-1 1.6E+0 2.9E+0 1.7E+0

1958 6.0E+0 8.3E+1 1.6E+1 9.1E-3 1.3E-3 6.6E-1 1.2E+0 4.1E+0 1.3E+0

1959 4.8E+0 1.8E+1 1.3E+1 7.3E-3 1.1E-3 5.4E-1 6.1E-1 3.4E+0 6.7E-1

1960 6.6E-2 2.6E+0 1.4E-1 2.5E-3 3.7E-4 6.6E-3 7.6E-2 2.1E-1 8.2E-2 

1961 4.6E-2 3.5E+0 3.3E-3 4.6E-4 6.8E-5 2.8E-3 2.0E-1 2.5E-1 2.1E-1

1962 2.1E-1 3.3E+0 4.6E-1 3.0E-4 4.5E-5 2.1E-2 2.0E-1 3.8E-1 2.2E-1 

1963 3.3E+0 2.1E+0 9.4E+0 2.8E-3 4.2E-4 3.7E-1 1.3E-1 2.8E+0 1.4E-1

1964 1.8E+0 1.1E-1 0.0E+0 1.1E-4 1.6E-5 2.8E+1 3.0E-2 7.1E-1 2.0E+0

1965 4.6E+0 7.4E-1 0.0E+0 4.7E-3 7.0E-4 2.0E+0 0.0E+0 2.7E+0 1.8E+0

1966 2.8E+0 4.3E-1 0.0E+0 1.1E-3 1.6E-4 1.2E+1 0.0E+0 7.8E-1 1.2E+0

1967 7.0E-2 1.8E-1 0.0E+0 1.2E-4 1.8E-5 1.7E+0 0.0E+0 2.1E-1 6.6E-1

1968 5.0E+0 3.4E-1 0.0E+0 2.3E-3 3.4E-4 7.4E-1 0.0E+0 1.2E+0 6.4E-1 

1969 2.8E-1 4.8E-1 0.0E+0 5.0E-4 7.4E-5 3.7E-1 0.0E+0 3.6E-1 5.3E-1

1970 6.6E-1 2.5E-5 0.0E+0 7.0E-4 1.1E-4 3.0E-1 0.0E+0 2.7E-1 5.5E-1 

1971 2.5E+0 7.0E-1 0.0E+0 2.1E-3 3.1E-4 3.2E+0 0.0E+0 1.1E+0 4.7E-1

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1 

1973 2.4E-2 1.4E-5 0.0E+0 2.4E-4 3.5E-5 1.1E-1 0.0E+0 6.2E-2 2.2E-2

1974 7.9E-3 1.3E-3 0.0E+0 1.1E-4 9.7E-6 4.4E-2 0.0E+0 3.7E-2 1.2E-1 

1975 3.8E-3 1.9E-3 0.0E+0 5.5E-5 1.1E-5 2.7E-2 0.0E+0 8.3E-3 1.1E-1

1976 7.4E-5 4.6E-5 0.0E+0 1.2E-5 5.1E-6 1.2E-3 0.0E+0 5.5E-4 4.3E-2 

1977 2.9E-5 2.0E-5 0.0E+0 1.1E-5 4.8E-6 1.5E-3 0.0E+0 7.2E-4 6.1E-2

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1 

1979 7.6E-5 4.2E-5 0.0E+0 2.0E-5 2.2E-6 5.5E-4 0.0E+0 3.8E-3 2.2E-2

1980 1.2E-4 6.2E-4 0.0E+0 1.3E-5 1.7E-6 2.7E-4 0.0E+0 1.8E-4 1.3E-1 

1981 1.2E-4 1.6E-3 0.0E+0 2.6E-6 4.8E-7 2.6E-3 0.0E+0 1.4E-4 8.6E-2

1982 6.6E-5 2.0E-5 0.0E+0 6.2E-6 6.9E-7 1.9E-4 0.0E+0 1.2E-4 3.1E-2

1983 1.2E-4 6.2E-4 0.0E+0 5.2E-5 6.9E-6 1.0E-3 0.0E+0 4.5E-5 1.5E-2

1984 1.2E-4 4.2E-5 0.0E+0 8.0E-6 3.2E-6 1.4E-4 0.0E+0 5.5E-5 6.2E-3

1985 1.2E-4 9.0E-4 0.0E+0 2.3E-6 4.5E-7 1.0E-3 0.0E+0 6.6E-5 9.9E-2

1986 1.2E-4 3.8E-5 0.0E+0 5.5E-7 4.2E-8 1.0E-3 0.0E+0 6.9E-6 1.9E-2

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1

1988 1.2E-4 5.9E-6 0.0E+0 4.8E-7 7.3E-8 6.6E-3 0.0E+0 1.2E-5 2.2E-1

1989 2.9E-4 9.7E-6 0.0E+0 5.7E-9 8.1E-10 1.6E-4 0.0E+0 7.4E-6 6.7E-2

1990 1.2E-4 3.8E-5 0.0E+0 9.4E-10 9.4E-10 4.2E-5 0.0E+0 2.0E-7 3.1E-2

1991 1.2E-4 1.6E-5 0.0E+0 1.0E-10 1.0E-10 5.2E-5 0.0E+0 9.7E-5 1.8E-2

1992 1.2E-4 1.8E-5 0.0E+0 1.7E-7 1.7E-7 1.4E-5 0.0E+0 8.3E-6 3.1E-2 

1993 0.0E+0 3.8E-6 0.0E+0 0.0E+0 8.7E-11 3.5E-5 0.0E+0 3.8E-5 0.0E+0

1994 0.0E+0 3.1E-5 0.0E+0 0.0E+0 4.6E-8 0.0E+0 0.0E+0 8.0E-5 0.0E+0

1995 0.0E+0 4.8E-5 0.0E+0 7.7E-8 1.3E-8 0.0E+0 0.0E+0 5.4E-6 0.0E+0

1996 0.0E+0 2.8E-5 0.0E+0 2.2E-7 4.5E-9 0.0E+0 0.0E+0 1.1E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 1.8E-7 5.5E-8 0.0E+0 0.0E+0 2.4E-5 0.0E+0

1998 0.0E+0 2.3E-5 0.0E+0 1.7E-7 1.8E-8 0.0E+0 0.0E+0 1.1E-5 0.0E+0

1999 0.0E+0 3.1E-5 0.0E+0 7.5E-8 7.1E-9 0.0E+0 0.0E+0 4.4E-6 0.0E+0

2000 0.0E+0 1.9E-2 0.0E+0 3.6E-4 3.6E-6 0.0E+0 0.0E+0 3.5E-2 0.0E+0

2001 0.0E+0 1.0E-2 0.0E+0 2.4E-6 2.9E-4 0.0E+0 0.0E+0 1.2E-3 0.0E+0

2002 0.0E+0 1.6E-4 0.0E+0 6.6E-6 1.8E-5 0.0E+0 0.0E+0 3.8E-3 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 33 of 40

Table 4-8. Intake (Bq yr-1) by Year for TRA 1952-2002 . 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 

240

Ru-106 Sr-89 Sr-90 Y-91

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 1.7E+0 2.9E+0 3.7E+0

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 4.3E+0 5.1E+0 9.2E+0

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 6.0E+0 7.3E+0 1.3E+1

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 7.2E+0 7.6E+0 1.5E+1

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 6.6E+0 1.2E+1 7.3E+0

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 5.1E+0 1.8E+1 5.6E+0

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 2.6E+0 1.5E+1 2.9E+0

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 3.2E-1 9.0E-1 3.5E-1 

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 8.5E-1 1.1E+0 9.2E-1

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 8.6E-1 1.6E+0 9.3E-1 

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 5.4E-1 1.2E+1 6.0E-1

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 1.3E-1 3.0E+0 8.5E+0

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0

1968 2.1E+1 1.4E+0 0.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0

1972 1.2E+0 1.2E+0 0.0E+0 2.8E-3 4.2E-4 2.1E+0 0.0E+0 1.2E+0 7.2E-1 

1973 5.7E-2 3.2E-5 0.0E+0 5.6E-4 8.1E-5 2.7E-1 0.0E+0 1.5E-1 5.1E-2

1974 2.4E-2 3.8E-3 0.0E+0 3.3E-4 2.9E-5 1.3E-1 0.0E+0 1.1E-1 3.7E-1 

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1

1976 2.2E-4 1.4E-4 0.0E+0 3.5E-5 1.5E-5 3.5E-3 0.0E+0 1.7E-3 1.3E-1 

1977 2.9E-4 2.0E-4 0.0E+0 1.1E-4 4.8E-5 1.5E-2 0.0E+0 7.2E-3 6.1E-1

1978 1.6E-3 8.7E-3 0.0E+0 3.2E-4 3.4E-5 2.4E-2 0.0E+0 8.3E-3 1.5E+0

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1 

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2

1984 2.9E-4 9.7E-5 0.0E+0 1.9E-5 7.4E-6 3.2E-4 0.0E+0 1.3E-4 1.4E-2 

1985 4.1E-4 3.0E-3 0.0E+0 7.6E-6 1.5E-6 3.3E-3 0.0E+0 2.2E-4 3.3E-1

1986 4.1E-4 1.3E-4 0.0E+0 1.8E-6 1.4E-7 3.3E-3 0.0E+0 2.3E-5 6.2E-2 

1987 1.2E-3 1.8E-4 0.0E+0 5.9E-6 9.0E-7 1.3E-4 0.0E+0 9.7E-5 3.3E+0

1988 1.2E-3 5.9E-5 0.0E+0 4.8E-6 7.3E-7 6.6E-2 0.0E+0 1.2E-4 2.2E+0

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2

1992 4.1E-4 6.0E-5 0.0E+0 5.7E-7 5.7E-7 4.6E-5 0.0E+0 2.8E-5 1.0E-1 

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0

1994 0.0E+0 4.4E-5 0.0E+0 0.0E+0 6.6E-8 0.0E+0 0.0E+0 1.1E-4 0.0E+0

1995 0.0E+0 4.8E-5 0.0E+0 7.7E-8 1.3E-8 0.0E+0 0.0E+0 5.4E-6 0.0E+0

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0

1999 0.0E+0 7.2E-5 0.0E+0 1.8E-7 1.7E-8 0.0E+0 0.0E+0 1.0E-5 0.0E+0

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0

2002 0.0E+0 1.6E-3 0.0E+0 6.7E-5 1.8E-4 0.0E+0 0.0E+0 3.8E-2 0.0E+0
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 34 of 40

Table 4-9. Comparison of Calculated Facility Intakes with Intakes from Environmental Monitoring 

Results . 

Year Activity

Type

Average Annual

Concentration

Annual Inhaled

Quan. (Bq)

Table 4 Inhaled

Quan. (Bq)

1963 β-γ 1.7E-11 μCi/cc 1510 1310a

Pu-239 6.0E-16 μCi/cc 0.05 0.014a

1968 α 0.0022 pCi/M3 0.18 0.01a

β 0.64 pCi/M3 56 337a

I-131 <0.08 pCi/M3 <7.1 1.4a

1969 α 0.023 pCi/M3 2 2.4E-3a

β 2.95 pCi/M3 262 118a

I-131 0.123 pCi/M3 10.9 2.1a

1970 gross β 6.0E-13 μCi/ml 53 74a

max. gr. Β @ CFA 8.1E-13 μCi/ml 72 74a

1973 gross β 95+42 fCi/M3 8.4 0.8a

EBR-I 90Sr 3.4 fCi/M3 0.3 0.15b

95Nb 1.0 fCi/M3 0.09 --- 137Cs 7-17 fCi/M3 0.6-1.5 --- 144Ce 4-8 fCi/M3 0.36-0.71 0.057b

EFS 90Sr 5.9 fCi/M3 0.52 0.15a

95Nb 0.9-2.4 fCi/M3 0.08-0.2 --- 106Ru 6-9.8 fCi/M3 0.53-0.87 0.27a

134Cs 0.8-1.6 fCi/M3 0.07-0.14 --- 137Cs 17-27 fCi/M3 1.5-2.4 ---

1976c gross β 3-6E-14 μCi/ml 2.5-5 0.6-25d

1986 85Kr @ CFA 3.7E-11 μCi/ml 3290 890e

1988 85Kr @ CFA 1.1E-10 μCi/ml 9770 14,000e

1990 85Kr @ CFA 2.7E-11 μCi/ml 2400 690e

a – values taken from Table 4-3 for CFA.

b – Values taken from Table 4-5 for RWMC since EBR-I is near RWMC.

c – of 90 monthly values (January through September) for 10 facility areas 89 values were between 3E-14 and 6E-14 μCi/ml.

d – using the current tables with 11 radionuclides the inhaled quantity is about 0.6 Bq; with the original tables with 44 

radionuclides the inhaled quantity is about 25 Bq.

e – values were derived from tables in an earlier version of the TBD report that contained concentrations of all INEEL 

released radionuclides.
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 35 of 40

Table 4-10: Intakes (Bq/event) for Criticalities that occurred at the INEEL. 

Date 10/16/1959 10/16/1959 1/25/1961 1/25/1961 10/17/1978 10/17/1978 11/18/1958

Event Criticality Criticality Criticality Criticality Criticality Criticality IET 13

Exposure 

Location

South 

ICPP

CFA Area South 

ICPP

CFA Area South 

ICPP

CFA Area LPTF/STPF

Rb-89 5.2E+5 8.8E+4 1.5E+1 1.0E+0 2.2E+3 3.7E+4 8.1E+2

Sr-91 5.4E+4 9.2E+3 1.7E+2 1.1E+1 1.7E+1 7.0E+2 4.1E+2

Sr-92 6.5E+4 1.1E+4 1.5E+2 9.7E+0 5.6E+0 2.1E+2 1.9E+2

Y-92 7.5E+3 1.3E+3 7.3E+1 4.9E+0 9.1E-2 2.6E+1 5.6E+1

Y-93 5.9E+3 1.0E+3 1.9E+1 1.3E+0 1.3E-1 1.5E+1 2.0E+1

Te-133 1.1E+3 8.0E+3

I-131(elem.) 2.9E+2 5.0E+1 1.9E+0 1.3E-1 3.9E-2 1.7E+0 3.4E+0

I-133 6.2E+3 1.1E+3 3.8E+1 2.5E+0 4.4E+0 5.8E+2 7.1E+1

I-134 4.6E+4 7.8E+3 7.2E+1 4.8E+0 1.1E+2 3.4E+3 2.0E+2

I-135 2.0E+4 3.4E+3 1.1E+2 7.2E+0 5.5E+1 2.2E+3 4.6E+2

Cs-138 7.7E+5 1.3E+5 5.2E+2 5.6E+4 5.0E+3

Ba-139 4.0E+5 6.8E+4 6.6E+2 4.4E+1 1.4E+2 1.8E+4 3.4E+3

La-141 2.8E+4 4.7E+3 1.0E+2 6.8E+0 4.6E-1 9.0E+1 2.0E+2

La-142 2.3E+4 3.9E+3 4.3E+1 2.8E+0 4.8E-1 5.7E+1 9.3E+1

U-234 7.1E-2 

Dose Fract 93.9% 92.5% 94.5% 94.5% 98.3% 97.1% 94.4%

Table 4-11: Intakes (Bq/event) for Special Tests at the INEEL . 

Date Event Exposure 

Location

Sr-89 Sr-90 Y-91 Zr-95 Ru-103 Ru-106 I-131

(elem.)

Ce-144 Pr-143

3/20/1957 FEBT A 

(14:19)

TAN 3.9E-2 6.1E-4 4.7E-2 5.0E-2 2.5E-2 1.1E-3 2.0E-2 1.8E-2 1.3E-2

3/20/1957 FEBT B 

(18:47)

TAN 1.3E-2 3.8E-3 2.4E-2 3.2E-2 3.8E-3 4.5E-3 9.0E-9 6.5E-2 2.6E-6

10/29/58 

10/30/58

FECF 

Filter 

Break

South 

ICPP

4.2E-2 4.2E-2 9.1E-2 1.3E-1 8.3E-3 4.3E-2 2.7E-13 5.9E-1 1.8E-7

10/29/58 

10/30/58

FECF 

Filter 

Break

CFA Area 1.8E+0 1.8E+0 3.9E+0 5.8E+0 3.6E-1 1.9E+0 1.2E-11 2.6E+1 7.6E-6
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 36 of 40

Table 4-12: Intakes (Bq/event) for Initial Engine Tests at the INEEL . 

Period Test

Exposure 

Location

Rb-89 Sr-89 I-131 

(elem.)

I-133 I-135 Cs-138 U-234

4/24 - 5/19/59 IET 14 TRA/ICPP 7.6E+0 2.3E-2 2.7E+0 1.7E+1 2.4E+1 5.5E+1 4.8E-6

4/24 - 5/19/59 IET 14 CFA 6.4E+0 2.0E-2 2.3E+0 1.4E+1 2.0E+1 4.6E+1 4.0E-6

6/16-6/24/59 IET 15(B) TRA/ICPP 4.3E-1 1.1E-3 5.9E-1 2.6E+0 4.2E+0 3.0E+0 5.1E-5

6/16-6/24/59 IET 15(B) CFA 3.6E-1 9.3E-4 5.0E-1 2.2E+0 3.6E+0 2.5E+0 4.3E-5

7/28 - 10/9/59 IET 16 SPERT 2.7E-2 3.6E-4 6.4E-4 1.3E-2 2.8E-2 9.7E-1 1.2E-7

7/28 - 10/9/59 IET 16 S. gate 2.2E-2 2.8E-4 5.1E-4 1.1E-2 2.2E-2 7.7E-1 9.7E-8

10/12-12/12/59 IET 17(B) TRA/ICPP 2.9E-3 8.2E-3 7.5E-1 3.2E+0 3.3E+0 8.2E-1 7.2E-6

10/12-12/12/59 IET 17(B) CFA 2.4E-3 6.9E-3 6.3E-1 2.7E+0 2.7E+0 6.9E-1 6.1E-6

1/26-2/7/60 IET 18 SPERT 1.2E-2 2.6E-3 8.3E+0 4.7E+1 3.1E+1 4.9E-1 8.1E-6

1/26-2/7/60 IET 18 S. gate 9.4E-3 2.0E-3 6.6E+0 3.7E+1 2.5E+1 3.9E-1 6.4E-6

2/17 - 2/29/60 IET 19(A) TRA/ICPP 1.7E-1 8.9E-3 1.3E+0 8.8E+0 1.3E+1 6.4E+0 5.5E-7

2/17 - 2/29/60 IET 19(A) CFA 1.4E-1 7.5E-3 1.1E+0 7.4E+0 1.1E+1 5.4E+0 4.6E-7

11/22-11/30/60 IET 25(A) TRA/ICPP 9.3E-4 6.1E-4 4.7E-1 5.2E+0 6.7E+0 2.0E-1 2.0E-6

11/22-11/30/60 IET 25(A) CFA 7.8E-4 5.2E-4 3.9E-1 4.3E+0 5.6E+0 1.7E-1 1.7E-6

12/1-12/15/60 IET 25(B) SPERT 1.2E-3 8.0E-4 1.4E+0 8.7E+0 8.3E+0 1.3E-1 3.2E-6

12/1-12/15/60 IET 25(B) S. gate 9.5E-4 6.4E-4 1.1E+0 6.9E+0 6.6E+0 1.1E-1 2.5E-6

12/23-12/28/60 IET 26(A) TRA/ICPP 2.2E+0 1.5E-2 2.4E+0 9.5E+0 1.4E+1 2.0E+1 3.3E-5

12/23-12/28/60 IET 26(A) CFA 1.9E+0 1.2E-2 2.0E+0 8.0E+0 1.1E+1 1.7E+1 2.8E-5
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 37 of 40

Table 4-13. INEEL facility fence direct gamma values (TLD – Background) (mR). 

Year ARA I & II SPERT TAN-TSF TAN-LOFT TAN-LPT CFA TRA ICPP RWMC EBR-II TREAT Backgrnd

1952-72 226 12 42 10 12 52 438 446 32 36 18 100-150

1973 86 21 41 17 12 53 306 405 32 37 19 121

1974 162 48 8 7 0 59 320 627 370 35 17 123

1975 114 16 29 11 7 17 195 357 265 32 8 118

1976 66 27 22 15 12 20 140 311 155 56 50 113

1977 41 5 0 4 4 1 137 318 189 22 0 132

1978 52 12 2 9 4 7 143 251 106 56 2 129

1979 63 18 10 17 7 14 159 236 65 59 5 113

1980 65 17 8 19 13 18 251 203 57 51 12 119

1981 63 18 8 17 10 14 231 255 42 28 9 118

1982 50 26 6 12 6 10 163 124 42 20 12 117

1983 78 17 23 26 19 18 174 141 50 24 10 115

1984 80 19 11 19 12 15 205 181 48 31 13 124

1985 80 19 11 19 12 15 205 181 48 31 13 124

1986 80 19 11 19 12 15 205 181 48 31 13 124

1987 80 19 11 19 12 15 205 181 48 31 13 124

1988 80 19 11 19 12 15 205 181 48 31 13 124

1989 80 19 11 19 12 15 205 181 48 31 13 124

1990 80 10 10 11 9 11 28 39 27 19 13 124

1991 80 10 10 11 9 11 28 39 27 19 13 124

1992 80 10 10 11 9 11 28 39 27 19 13 124

1993 77 19 18 23 15 15 48 37 24 28 16 111

1994 69 0 0 0 0 0 24 28 25 15 3 130

1995 91 6 4 8 2 11 31 43 42 17 7 116

1996 52 4 14 0 0 13 28 49 40 22 21 129

1997 46 10 3 8 0 9 29 44 17 16 16 128

1998 62 8 0 5 0 12 25 31 20 0 11 131

1999 49 13 0 0 0 5 10 38 22 13 13 122

2000 28 16 7 16 8 19 40 55 61 25 26 129

2001 31 3 0 0 0 0 27 32 25 0 3 140

2002 41 11 0 0 0 9 34 54 33 18 39 120
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-4 Page 38 of 40

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